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Note: This page contains sample records for the topic "uranium enrichment decontamination" from the National Library of EnergyBeta (NLEBeta).
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1

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...

2

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

3

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

4

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...

5

Uranium enrichment decontamination and decommissioning fund, 1995 report  

SciTech Connect

This report describes strategies for the decontamination and decommissioning of gaseous diffusion plants. Progress in remedial action activities are discussed.

1996-11-01T23:59:59.000Z

6

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

7

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

8

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

9

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

10

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

11

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

12

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%.

13

DECONTAMINATION OF URANIUM  

DOE Patents (OSTI)

This patent deals with the separation of rare earth and other fission products from neutron bombarded uranium. This is accomplished by melting the uranium in contact with either thorium oxide, maguesium oxide, alumnum oxide, beryllium oxide, or uranium dioxide. The melting is preferably carried out at from 1150 deg to 1400 deg C in an inert atmosphere, such as argon or helium. During this treatment a scale of uranium dioxide forms on the uranium whtch contains most of the fission products.

Feder, H.M.; Chellew, N.R.

1958-02-01T23:59:59.000Z

14

DECONTAMINATION OF URANIUM  

DOE Patents (OSTI)

A process is given for separating fission products from uranium by extracting the former into molten aluminum. Phase isolation can be accomplished by selectively hydriding the uranium at between 200 and 300 deg C and separating the hydride powder from coarse particles of fissionproduct-containing aluminum. (AEC)

Spedding, F.H.; Butler, T.A.

1962-05-15T23:59:59.000Z

15

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...

16

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...

17

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...

18

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

19

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

20

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...

Note: This page contains sample records for the topic "uranium enrichment decontamination" 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

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

22

Plutonium Decontamination of Uranium using CO2 Cleaning  

SciTech Connect

A concern of the Department of Energy (DOE) Environmental Management (EM) and Defense Programs (DP), and of the Los Alamos National Laboratory (LANL) and the Lawrence Livermore National Laboratory (LLNL), is the disposition of thousands of legacy and recently generated plutonium (Pu)-contaminated, highly enriched uranium (HEU) parts. These parts take up needed vault space. This presents a serious problem for LLNL, as site limit could result in the stoppage of future weapons work. The Office of Fissile Materials Disposition (NN-60) will also face a similar problem as thousands of HEU parts will be created with the disassembly of site-return pits for plutonium recovery when the Pit Disassembly and Conversion Facility (PDCF) at the Savannah River Site (SRS) becomes operational. To send HEU to the Oak Ridge National Laboratory and the Y-12 Plant for disposition, the contamination for metal must be less than 20 disintegrations per minute (dpm) of swipable transuranic per 100 cm{sup 2} of surface area or the Pu bulk contamination for oxide must be less than 210 parts per billion (ppb). LANL has used the electrolytic process on Pu-contaminated HEU weapon parts with some success. However, this process requires that a different fixture be used for every configuration; each fixture cost approximately $10K. Moreover, electrolytic decontamination leaches the uranium metal substrate (no uranium or plutonium oxide) from the HEU part. The leaching rate at the uranium metal grain boundaries is higher than that of the grains and depends on the thickness of the uranium oxide layer. As the leaching liquid flows past the HEU part, it carries away plutonium oxide contamination and uranium oxide. The uneven uranium metal surface created by the leaching becomes a trap for plutonium oxide contamination. In addition, other DOE sites have used CO{sub 2} cleaning for Pu decontamination successfully. In the 1990's, the Idaho National Engineering Laboratory investigated this technology and showed that CO{sub 2} pellet blasting (or CO{sub 2} cleaning) reduced both fixed and smearable contamination on tools. In 1997, LLNL proved that even tritium contamination could be removed from a variety of different matrices using CO{sub 2}cleaning. CO{sub 2} cleaning is a non-toxic, nonconductive, nonabrasive decontamination process whose primary cleaning mechanisms are: (1) Impact of the CO{sub 2} pellets loosens the bond between the contaminant and the substrate. (2) CO{sub 2} pellets shatter and sublimate into a gaseous state with large expansion ({approx}800 times). The expanding CO{sub 2} gas forms a layer between the contaminant and the substrate that acts as a spatula and peels off the contaminant. (3) Cooling of the contaminant assists in breaking its bond with the substrate. Thus, LLNL conducted feasibility testing to determine if CO{sub 2} pellet blasting could remove Pu contamination (e.g., uranium oxide) from uranium metal without abrading the metal matrix. This report contains a summary of events and the results of this test.

Blau, M

2002-12-01T23:59:59.000Z

23

Plutonium Decontamination of Uranium using CO2 Cleaning  

SciTech Connect

A concern of the Department of Energy (DOE) Environmental Management (EM) and Defense Programs (DP), and of the Los Alamos National Laboratory (LANL) and the Lawrence Livermore National Laboratory (LLNL), is the disposition of thousands of legacy and recently generated plutonium (Pu)-contaminated, highly enriched uranium (HEU) parts. These parts take up needed vault space. This presents a serious problem for LLNL, as site limit could result in the stoppage of future weapons work. The Office of Fissile Materials Disposition (NN-60) will also face a similar problem as thousands of HEU parts will be created with the disassembly of site-return pits for plutonium recovery when the Pit Disassembly and Conversion Facility (PDCF) at the Savannah River Site (SRS) becomes operational. To send HEU to the Oak Ridge National Laboratory and the Y-12 Plant for disposition, the contamination for metal must be less than 20 disintegrations per minute (dpm) of swipable transuranic per 100 cm{sup 2} of surface area or the Pu bulk contamination for oxide must be less than 210 parts per billion (ppb). LANL has used the electrolytic process on Pu-contaminated HEU weapon parts with some success. However, this process requires that a different fixture be used for every configuration; each fixture cost approximately $10K. Moreover, electrolytic decontamination leaches the uranium metal substrate (no uranium or plutonium oxide) from the HEU part. The leaching rate at the uranium metal grain boundaries is higher than that of the grains and depends on the thickness of the uranium oxide layer. As the leaching liquid flows past the HEU part, it carries away plutonium oxide contamination and uranium oxide. The uneven uranium metal surface created by the leaching becomes a trap for plutonium oxide contamination. In addition, other DOE sites have used CO{sub 2} cleaning for Pu decontamination successfully. In the 1990's, the Idaho National Engineering Laboratory investigated this technology and showed that CO{sub 2} pellet blasting (or CO{sub 2} cleaning) reduced both fixed and smearable contamination on tools. In 1997, LLNL proved that even tritium contamination could be removed from a variety of different matrices using CO{sub 2}cleaning. CO{sub 2} cleaning is a non-toxic, nonconductive, nonabrasive decontamination process whose primary cleaning mechanisms are: (1) Impact of the CO{sub 2} pellets loosens the bond between the contaminant and the substrate. (2) CO{sub 2} pellets shatter and sublimate into a gaseous state with large expansion ({approx}800 times). The expanding CO{sub 2} gas forms a layer between the contaminant and the substrate that acts as a spatula and peels off the contaminant. (3) Cooling of the contaminant assists in breaking its bond with the substrate. Thus, LLNL conducted feasibility testing to determine if CO{sub 2} pellet blasting could remove Pu contamination (e.g., uranium oxide) from uranium metal without abrading the metal matrix. This report contains a summary of events and the results of this test.

Blau, M

2002-12-01T23:59:59.000Z

24

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...

25

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...

26

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...

27

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...

28

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...

29

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...

30

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...

31

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...

32

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...

33

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...

34

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...

35

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

36

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

37

ELECTROCHEMICAL DECONTAMINATION AND RECOVERY OF URANIUM VALUES  

DOE Patents (OSTI)

An electrochemical process is described for separating uranium from fission products. The method comprises subjecting the mass of uranium to anodic dissolution in an electrolytic cell containing aqueous alkali bicarbonate solution as its electrolyte, thereby promoting a settling from the solution of a solid sludge from about the electrodes and separating the resulting electrolyte solution containing the anodically dissolved uranium from the sludge which contains the rare earth fission products.

McLaren, J.A.; Goode, J.H.

1958-05-13T23:59:59.000Z

38

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)

39

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

40

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...

Note: This page contains sample records for the topic "uranium enrichment decontamination" 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

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...

42

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

43

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

44

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

45

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

46

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

47

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

48

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

49

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

50

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

51

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

52

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

53

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

54

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...

55

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

56

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

57

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,...

58

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...

59

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...

60

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...

Note: This page contains sample records for the topic "uranium enrichment decontamination" 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

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...

62

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...

63

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

64

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

65

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

66

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

67

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

68

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

69

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

70

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

71

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

72

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

73

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

74

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

75

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

76

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

77

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

78

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

79

Surface Decontamination of System Components in Uranium Conversion Plant at KAERI  

SciTech Connect

A chemical decontamination process using nitric acid solution was selected as in-situ technology for recycle or release with authorization of a large amount of metallic waste including process system components such as tanks, piping, etc., which is generated by dismantling a retired uranium conversion plant at Korea Atomic Energy Research Institute (KAERI). The applicability of nitric acid solution for surface decontamination of metallic wastes contaminated with uranium compounds was evaluated through the basic research on the dissolution of UO2 and ammonium uranyl carbonate (AUC) powder. Decontamination performance was verified by using the specimens contaminated with such uranium compounds as UO2 and AUC taken from the uranium conversion plant. Dissolution rate of UO2 powder is notably enhanced by the addition of H2O2 as an oxidant even in the condition of a low concentration of nitric acid and low temperature compared with those in a nitric acid solution without H2O2. AUC powders dissolve easily in nitric acid solutions until the solution pH attains about 2.5 {approx} 3. Above that solution pH, however, the uranium concentration in the solution is lowered drastically by precipitation as a form of U3(NH3)4O9 . 5H2O. Decontamination performance tests for the specimens contaminated with UO2 and AUC were quite successful with the application of decontamination conditions obtained through the basic studies on the dissolution of UO2 and AUC powders.

Choi, W. K.; Kim, K. N.; Won, H. J.; Jung, C. H.; Oh, W. Z.

2003-02-25T23:59:59.000Z

80

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 enrichment decontamination" 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

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

82

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

83

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

84

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.

85

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...

86

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)

87

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.

88

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

89

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

90

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

91

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

92

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

93

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

94

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

95

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

96

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

97

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

98

An integrated approach to the characterization and decontamination of uranium contaminated soils  

SciTech Connect

An Integrated Demonstration (ID) Program, hosted by the Fernald Environmental Restoration Management Company, has been established for investigating technologies applicable to the characterization and remediation of soils contaminated with uranium. Chemical and physical characterization of Fernald soils and the uranium wastes contained therein is being accomplished by means of standard analytical techniques as well as a suite of non-standard microscopy and spectroscopy techniques. Likewise, a suite of physical and chemical extraction technologies are being designed and tested for accomplishing soil decontamination. However, the main theme of this paper is not the technologies being tested but the approach taken to integrate characterization, decontamination, and risk assessment efforts. It is the authors intent to outline the critical components of an integrated approach for characterizing and remediating uranium contaminated soils as well as provide a real-world example based on the lessons learned in the ID program.

Tidwell, V. [Sandia National Labs., Albuquerque, NM (United States); Francis, C.; Armstrong, A. [Oak Ridge National Lab., TN (United States); Dyer, R. [Environmental Protection Agency, Washington, DC (United States)

1994-02-01T23:59:59.000Z

99

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

100

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 enrichment decontamination" 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

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

102

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

103

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

104

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

105

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

106

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

107

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

108

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

109

PROCESS FOR DECONTAMINATING THORIUM AND URANIUM WITH RESPECT TO RUTHENIUM  

DOE Patents (OSTI)

The control of ruthenium extraction in solvent-extraction processing of neutron-irradiated thorium is presented. Ruthenium is rendered organic-insoluble by the provision of sulfite or bisulfite ions in the aqueous feed solution. As a result the ruthenium remains in the aqueous phase along with other fission product and protactinium values, thorium and uranium values being extracted into the organic phase. This process is particularly applicable to the use of a nitrate-ion-deficient aqueous feed solution and to the use of tributyl phosphate as the organic extractant.

Meservey, A.A.; Rainey, R.H.

1959-10-20T23:59:59.000Z

110

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

111

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

112

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

113

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

114

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

115

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

116

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

117

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

118

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. NNSAs 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 capabilitiesin this case, by automating the monitoring of uranium enrichment in the entire inventory of a fuel fabrication facility. One such system is HEVAhybrid 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

119

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

120

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

Note: This page contains sample records for the topic "uranium enrichment decontamination" 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

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

122

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

123

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

124

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

125

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

126

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

127

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

128

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

129

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

130

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

131

Uranium polishing of plutonium contaminated sludge in support of the highly enriched uranium decontamination effort  

SciTech Connect

ARIES is the prototype development and demonstration system for government furnished Pit Disassembly and Conversion Facility technologies. ARIES separates a nuclear weapon's Pu from other components in a pit. AIRES prepares weapons Pu for long-term storage and disposition in a form quantifiably verified by nondestructive assay. Assay results can be presented for international inspection and safeguards.

Cisneros, M. R. (Michael R.); Costa, D. A. (David A.); Montoya, W. E. (Willie E.)

2004-01-01T23:59:59.000Z

132

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

133

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.

134

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...

135

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

136

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

137

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...

138

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

139

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

Note: This page contains sample records for the topic "uranium enrichment decontamination" 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

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

142

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

143

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

144

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

145

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

146

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

147

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

148

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

149

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

150

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

151

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

152

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

153

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

154

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

155

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

156

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

157

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.

158

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

159

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

160

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

Note: This page contains sample records for the topic "uranium enrichment decontamination" 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

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

162

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

163

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

164

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

165

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

166

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

167

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

168

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

169

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

170

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

171

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

172

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

173

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

174

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 fuelâ??s complete lifetime. An assessment of the fuelâ??s viability for use with the Advanced Fuel Cycle Initiativeâ??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

175

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

176

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

177

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

178

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,...

179

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

180

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

Note: This page contains sample records for the topic "uranium enrichment decontamination" 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

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

182

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

183

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

184

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

185

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

186

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

187

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 MissouriColumbia, 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

188

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

189

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

190

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

191

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 worlds 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

192

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

193

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

194

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

195

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

196

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

197

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

198

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

199

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

200

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

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201

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

202

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

203

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 facilitys 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

204

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 samples 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

205

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

206

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

207

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

208

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

209

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

210

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

211

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

212

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

213

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

214

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 eff. 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

215

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

216

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

217

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

218

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

219

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

220

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

Note: This page contains sample records for the topic "uranium enrichment decontamination" from the National Library of EnergyBeta (NLEBeta).
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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.


221

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

222

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

223

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

224

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

225

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

226

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

227

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

228

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

229

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 .......................................................................

230

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

231

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

232

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

233

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

234

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

SciTech Connect

In June 2009 Romania successfully completed the worlds 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

235

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

236

Environmental decontamination  

SciTech Connect

The record of the proceedings of the workshop on environmental decontamination contains twenty-seven presentations. Emphasis is placed upon soil and surface decontamination, the decommissioning of nuclear facilities, and assessments of instrumentation and equipment used in decontamination. (DLS)

Cristy, G.A.; Jernigan, H.C. (eds.)

1981-02-01T23:59:59.000Z

237

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

238

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

239

PYROCHEMICAL DECONTAMINATION METHOD FOR REACTOR FUEL  

DOE Patents (OSTI)

A pyro-chemical method is presented for decontaminating neutron irradiated uranium and separating plutonium therefrom by contact in the molten state with a metal chloride salt. Uranium trichloride and uranium tetrachloride either alone or in admixture with alkaline metal and alkaline eanth metal fluorides under specified temperature and specified phase ratio conditions extract substantially all of the uranium from the irradiated uranium fuel together with certain fission products. The phases are then separated leaving purified uranium metal. The uranium and plutonium in the salt phase can be reduced to forin a highly decontaminated uraniumplutonium alloy. The present method possesses advantages for economically decontaminating irradiated nuclear fuel elements since irradiated fuel may be proccessed immediately after withdrawal from the reactor and the uranium need not be dissolved and later reduced to the metallic form. Accordingly, the uranium may be economically refabricated and reinserted into the reactor.

Buyers, A.G.

1959-06-30T23:59:59.000Z

240

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

Note: This page contains sample records for the topic "uranium enrichment decontamination" 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

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 ...

242

Decontamination Handbook  

Science Conference Proceedings (OSTI)

Decontamination has played a key role in reducing collective exposures at U.S. nuclear power plants over the last decade. Today's challenges are to decrease costs and minimize the impact of decontamination activities on outage duration. This handbook compiles lessons learned from sub- and full-system decontaminations as well as decontamination after plant shutdown. It provides critical information about dilute solvents, corrosion issues, the effect of coolant chemistry on decontamination effectiveness, r...

1999-07-14T23:59:59.000Z

243

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

244

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

245

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

246

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

247

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

248

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

249

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

250

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

251

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

252

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

253

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

254

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

255

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

256

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

257

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

258

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

259

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

260

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

Note: This page contains sample records for the topic "uranium enrichment decontamination" 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

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

262

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...

263

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

264

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

265

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

266

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

267

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 Re, 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

268

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

269

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

270

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

271

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

272

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

273

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

274

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

SciTech Connect

In October of 1999, Lawrence Livermore National Laboratory (LLNL) began an effort to calibrate the LLNL passive-active neutron (PAN) drum shuffler for measurement of highly enriched uranium (HEU) oxide. A single unit of certified reference material (CRM) 149 [Uranium (93% Enriched) Oxide - U{sub 3}O{sub 8} Standard for Neutron Counting Measurements] was used to (1) develop a mass calibration curve for HEU oxide in the nominal range of 393 g to 3144 g {sup 235}U, and (2) perform a detailed axial and radial mapping of the detector response over a wide region of the PAN shuffler counting chamber. Results from these efforts were reported at the Institute of Nuclear Materials Management 41st Annual Meeting in July 2000. This paper describes subsequent efforts by LLNL to use a unit of CRM 146 [Uranium Isotopic Standard for Gamma Spectrometry Measurements] in consort with Monte Carlo simulations of the PAN shuffler response to CRM 149 and CRM 146 units and a selected set of containers with CRM 149-equivalent U{sub 3}O{sub 8} to (1) extend the low range of the reported mass calibration curve to 10 g {sup 235}U, (2) evaluate the effect of U{sub 3}O{sub 8} density (2.4 g/cm{sup 3} to 4.8 g/cm{sup 3}) and container size (5.24 cm to 12.17 cm inside diameter and 6.35 cm to 17.72 cm inside height) on the PAN shuffler response, and (3) develop mass calibration curves for U{sub 3}O{sub 8} enriched to 20.1 wt% {sup 235}U and 52.5 wt% {sup 235}U.

Mount, M; O' Connell, W; Cochran, C; Rinard, P; Dearborn, D; Endres, E

2002-05-17T23:59:59.000Z

275

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

SciTech Connect

As a follow-on to the Lawrence Livermore National Laboratory (LLNL) effort to calibrate the LLNL passive-active neutron drum (PAN) shuffler for measurement of highly enriched uranium (HEU) oxide, a method has been developed to extend the use of the PAN shuffler to the measurement of HEU in mixed uranium-plutonium (U-Pu) oxide. This method uses the current LLNL HEU oxide calibration algorithms, appropriately corrected for the mixed U-Pu oxide assay time, and recently developed PuO{sub 2} calibration algorithms to yield the mass of {sup 235}U present via differences between the expected count rate for the PuO{sub 2} and the measured count rate of the mixed U-Pu oxide. This paper describes the LLNL effort to use PAN shuffler measurements of units of certified reference material (CRM) 149 [uranium (93% Enriched) Oxide - U{sub 3}O{sub 8} Standard for Neutron Counting Measurements] and CRM 146 [Uranium Isotopic Standard for Gamma Spectrometry Measurements] and a selected set of LLNL PuO{sub 2}-bearing containers in consort with Monte Carlo simulations of the PAN shuffler response to each to (1) establish and validate a correction to the HEU calibration algorithm for the mixed U-Pu oxide assay time, (2) develop a PuO{sub 2} calibration algorithm that includes the effect of PuO{sub 2} density (2.4 g/cm{sup 3} to 4.8 g/cm{sup 3}) and container size (8.57 cm to 9.88 cm inside diameter and 9.60 cm to 13.29 cm inside height) on the PAN shuffler response, and (3) develop and validate the method for establishing the mass of {sup 235}U present in an unknown of mixed U-Pu oxide.

Mount, M; O' Connell, W; Cochran, C; Rinard, P; Dearborn, D; Endres, E

2002-05-23T23:59:59.000Z

276

MCNP-DSP calculations of the {sup 252}Cf-source-driven noise analysis measurements of highly enriched uranium metal cylinders  

SciTech Connect

This paper presents calculations of the {sup 252}Cf-source-driven noise analysis measurements for subcritical highly enriched uranium metal cylinders using the Monte Carlo code MCNP-DSP. This code directly calculates the noise analysis data from the {sup 252}Cf- source-driven noise analysis method for both neutron and gamma ray detectors. Direct calculation of experimental observables by the Monte Carlo method allows for the benchmarking of the calculational model and the cross sections and for determining the bias in the calculation.

Valentine, T.E.; Mihalczo, J.T.

1995-07-01T23:59:59.000Z

277

300 AREA URANIUM CONTAMINATION  

SciTech Connect

{sm_bullet} Uranium fuel production {sm_bullet} Test reactor and separations experiments {sm_bullet} Animal and radiobiology experiments conducted at the. 331 Laboratory Complex {sm_bullet} .Deactivation, decontamination, decommissioning,. and demolition of 300 Area facilities

BORGHESE JV

2009-07-02T23:59:59.000Z

278

IMPROVED PROCESSES FOR RECOVERING AND PURIFYING URANIUM  

DOE Patents (OSTI)

A process is described for reclaiming metallic uranium enriched with uranium-235 from the collector of a calutron upon which the enriched metallic uranium is Editor please delete 22166.

Price, T.D.; Henrickson, A.V.

1959-05-12T23:59:59.000Z

279

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

280

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...

Note: This page contains sample records for the topic "uranium enrichment decontamination" 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

Y-12 uranium storage facility?a dream come true?  

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

ranks and actually provides the first impedance for the just finished highly enriched uranium storage facility. Recently the Highly Enriched Uranium Material Facility was...

282

EIS-0360: Depleted Uranium Oxide Conversion Product at the Portsmouth, Ohio  

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

60: Depleted Uranium Oxide Conversion Product at the 60: Depleted Uranium Oxide Conversion Product at the Portsmouth, Ohio Site EIS-0360: Depleted Uranium Oxide Conversion Product at the Portsmouth, Ohio Site Summary This site-specific EIS analyzes the construction, operation, maintenance, and decontamination and decommissioning of the proposed depleted uranium hexafluoride (DUF6) conversion facility at three alternative locations within the Portsmouth site; transportation of all cylinders (DUF6, enriched, and empty) currently stored at the East Tennessee Technology Park (ETTP) near Oak Ridge, Tennessee, to Portsmouth; construction of a new cylinder storage yard at Portsmouth (if required) for ETTP cylinders; transportation of depleted uranium conversion products and waste materials to a disposal facility; transportation and sale of the hydrogen fluoride

283

Results from a "Proof-of-Concept" Demonstration of RF-Based Tracking of UF6 Cylinders during a Processing Operation at a Uranium Enrichment Plant  

SciTech Connect

Approved industry-standard cylinders are used globally for processing, storing, and transporting uranium hexafluoride (UF{sub 6}) at uranium enrichment plants. To ensure that cylinder movements at enrichment facilities occur as declared, the International Atomic Energy Agency (IAEA) must conduct time-consuming periodic physical inspections to validate facility records, cylinder identity, and containment. By using a robust system design that includes the capability for real-time unattended monitoring (of cylinder movements), site-specific rules-based event detection algorithms, and the capability to integrate with other types of monitoring technologies, one can build a system that will improve overall inspector effectiveness. This type of monitoring system can provide timely detection of safeguard events that could be used to ensure more timely and appropriate responses by the IAEA. It also could reduce reliance on facility records and have the additional benefit of enhancing domestic safeguards at the installed facilities. This paper will discuss the installation and evaluation of a radio-frequency- (RF-) based cylinder tracking system that was installed at a United States Enrichment Corporation Centrifuge Facility. This system was installed primarily to evaluate the feasibility of using RF technology at a site and the operational durability of the components under harsh processing conditions. The installation included a basic system that is designed to support layering with other safeguard system technologies and that applies fundamental rules-based event processing methodologies. This paper will discuss the fundamental elements of the system design, the results from this site installation, and future efforts needed to make this technology ready for IAEA consideration.

Pickett, Chris A [ORNL; Kovacic, Donald N [ORNL; Whitaker, J Michael [ORNL; Younkin, James R [ORNL; Hines, Jairus B [ORNL; Laughter, Mark D [ORNL; Morgan, Jim [Innovative Solutions; Carrick, Bernie [USEC; Boyer, Brian [Los Alamos National Laboratory (LANL); Whittle, K. [USEC

2008-01-01T23:59:59.000Z

284

Progress in developing processes for converting {sup 99}Mo production from high- to low-enriched uranium--1998.  

SciTech Connect

During 1998, the emphasis of our activities was focused mainly on target fabrication. Successful conversion requires a reliable irradiation target; the target being developed uses thin foils of uranium metal, which can be removed from the target hardware for dissolution and processing. This paper describes successes in (1) improving our method for heat-treating the uranium foil to produce a random-small grain structure, (2) improving electrodeposition of zinc and nickel fission-fragment barriers onto the foil, and (3) showing that these fission fragment barriers should be stable during transport of the targets following irradiation. A method was also developed for quantitatively electrodepositing uranium and plutonium contaminants in the {sup 99}Mo. Progress was also made in broadening international cooperation in our development activities.

Conner, C.

1998-10-28T23:59:59.000Z

285

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...

286

In-Situ Measurements of Low Enrichment Uranium Holdup Process Gas Piping at K-25 - Paper for Waste Management Symposia 2010 East Tennessee Technology Park Oak Ridge, Tennessee  

SciTech Connect

This document is the final version of a paper submitted to the Waste Management Symposia, Phoenix, 2010, abstract BJC/OR-3280. The primary document from which this paper was condensed is In-Situ Measurement of Low Enrichment Uranium Holdup in Process Gas Piping at K-25 Using NaI/HMS4 Gamma Detection Systems, BJC/OR-3355. This work explores the sufficiency and limitations of the Holdup Measurement System 4 (HJVIS4) software algorithms applied to measurements of low enriched uranium holdup in gaseous diffusion process gas piping. HMS4 has been used extensively during the decommissioning and demolition project of the K-25 building for U-235 holdup quantification. The HMS4 software is an integral part of one of the primary nondestructive assay (NDA) systems which was successfully tested and qualified for holdup deposit quantification in the process gas piping of the K-25 building. The initial qualification focused on the measurement of highly enriched UO{sub 2}F{sub 2} deposits. The purpose of this work was to determine if that qualification could be extended to include the quantification of holdup in UO{sub 2}F{sub 2} deposits of lower enrichment. Sample field data are presented to provide evidence in support of the theoretical foundation. The HMS4 algorithms were investigated in detail and found to sufficiently compensate for UO{sub 2}F{sub 2} source self-attenuation effects, over the range of expected enrichment (4-40%), in the North and East Wings of the K-25 building. The limitations of the HMS4 algorithms were explored for a described set of conditions with respect to area source measurements of low enriched UO{sub 2}F{sub 2} deposits when used in conjunction with a 1 inch by 1/2 inch sodium iodide (NaI) scintillation detector. The theoretical limitations of HMS4, based on the expected conditions in the process gas system of the K-25 building, are related back to the required data quality objectives (DQO) for the NBA measurement system established for the K-25 demolition project. The combined review of the HMS software algorithms and supporting field measurements lead to the conclusion that the majority of process gas pipe measurements are adequately corrected for source self-attenuation using HMS4. While there will be instances where the UO{sub 2}F{sub 2} holdup mass presents an infinitely thick deposit to the NaI-HMS4 system these situations are expected to be infrequent. This work confirms that the HMS4 system can quantify UO{sub 2}F{sub 2} holdup, in its current configuration (deposition, enrichment, and geometry), below the DQO levels for the K-25 building decommissioning and demolition project. For an area measurement of process gas pipe in the K-25 building, if an infinitely thick UO{sub 2}F{sub 2} deposit is identified in the range of enrichment of {approx}4-40%, the holdup quantity exceeds the corresponding DQO established for the K-25 building demolition project.

Rasmussen B.

2010-01-01T23:59:59.000Z

287

Radiometric Determination of Uranium in Natural Waters after Enrichment and Separation by Cation-Exchange and Liquid-Liquid Extraction  

E-Print Network (OSTI)

The alpha-radiometric determination of uranium after its pre-concentration from natural water samples using the cation-exchange resin Chelex-100, its selective extraction by tributylphosphate and electrodeposition on stainless steel discs is reported. The validity of the separation procedure and the chemical recoveries were checked by addition of uranium standard solution as well as by tracing with U-232. The average uranium yield was determined to be (97 +- 2) % for the cation-exchange, (95 +- 2) % for the liquid-liquid extraction, and more than 99% for the electrodeposition. Employing high-resolution alpha-spectroscopy, the measured activity of the U-238 and U-234 radioisotopes was found to be of similar magnitude; i.e. ~7 mBq/L and ~35 mBq/L for ground- and seawater samples, respectively. The energy resolution (FWHM) of the alpha-peaks was 22 keV, while the Minimum Detectable Activity (MDA) was estimated to be 1 mBq/L (at the 95% confidence limit).

I. Pashalidis; H. Tsertos

2003-04-28T23:59:59.000Z

288

Advanced technologies for decontamination and conversion of scrap metals  

Science Conference Proceedings (OSTI)

Recycle of radioactive scrap metals (RSM) from decommissioning of DOE uranium enrichment and nuclear weapons manufacturing facilities is mandatory to recapture the value of these metals and avoid the high cost of disposal by burial. The scrap metals conversion project detailed below focuses on the contaminated nickel associated with the gaseous diffusion plants. Stainless steel can be produced in MSC`s vacuum induction melting process (VIM) to the S30400 specification using nickel as an alloy constituent. Further the case alloy can be rolled in MSC`s rolling mill to the mechanical property specification for S30400 demonstrating the capability to manufacture the contaminated nickel into valuable end products at a facility licensed to handle radioactive materials. Bulk removal of Technetium from scrap nickel is theoretically possible in a reasonable length of time with the high calcium fluoride flux, however the need for the high temperature creates a practical problem due to flux volatility. Bulk decontamination is possible and perhaps more desirable if nickel is alloyed with copper to lower the melting point of the alloy allowing the use of the high calcium fluoride flux. Slag decontamination processes have been suggested which have been proven technically viable at the Colorado School of Mines.

Muth, T.R. [Manufacturing Sciences Corp., Oak Ridge, TN (United States); Moore, J.; Olson, D.; Mishra, B. [Colorado School of Mines, Golden, CO (United States)

1994-12-31T23:59:59.000Z

289

Depleted Uranium Health Effects  

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

Depleted Uranium Health Effects Depleted Uranium Health Effects Depleted Uranium line line Uranium Enrichment Depleted Uranium Health Effects Depleted Uranium Health Effects Discussion of health effects of external exposure, ingestion, and inhalation of depleted uranium. Depleted uranium is not a significant health hazard unless it is taken into the body. External exposure to radiation from depleted uranium is generally not a major concern because the alpha particles emitted by its isotopes travel only a few centimeters in air or can be stopped by a sheet of paper. Also, the uranium-235 that remains in depleted uranium emits only a small amount of low-energy gamma radiation. However, if allowed to enter the body, depleted uranium, like natural uranium, has the potential for both chemical and radiological toxicity with the two important target organs

290

Instrument calibration and measurement plan for the poorly measured/unmeasured category of highly enriched uranium at Lawrence Livermore National Laboratory  

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) metal and oxide items in its inventory. Because of a lack of traceable HEU standards, no items were deemed to fit the category of well measured. A subsequent DOE-HQ sponsored survey by New Brunswick Laboratory resulted in their preparation of a set of certified reference material (CRM) standards for HEU oxide (U{sub 3}O{sub 8}) that are projected for delivery during September of 1999. However, CRM standards for HEU metal are neither in preparation nor are they expected to be prepared within the foreseeable future. Consequently, HEU metal working standards must be developed if the poorly measured/unmeasured portion of the LLNL inventory is to be reclassified. This paper describes the approach that LLNL will take to (1) develop a set of HEU metal working standards; (2) develop HEU metal and oxide calibration curves for the passive-active neutron (PAN) shuffler that are functions of mass, enrichment, size, and shape; and (3) reclassify the poorly measured/unmeasured inventory through direct measurement or reprocessing of previously archived data.

Glosup, J; Mount, M E

1999-07-01T23:59:59.000Z

291

Uranium Purchases Report  

Reports and Publications (EIA)

Final issue. This report details natural and enriched uranium purchases as reported by owners and operators of commercial nuclear power plants. 1996 represents the most recent publication year.

Douglas Bonnar

1996-06-01T23:59:59.000Z

292

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 Re, 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 Re 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

293

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

294

Y-12 Plant Decontamination and Decommissioning Technology Logic Diagram for Building 9201-4. Volume 1: Technology evaluation  

Science Conference Proceedings (OSTI)

During World War 11, the Oak Ridge Y-12 Plant was built as part of the Manhattan Project to supply enriched uranium for weapons production. In 1945, Building 9201-4 (Alpha-4) was originally used to house a uranium isotope separation process based on electromagnetic separation technology. With the startup of the Oak Ridge K-25 Site gaseous diffusion plant In 1947, Alpha-4 was placed on standby. In 1953, the uranium enrichment process was removed, and installation of equipment for the Colex process began. The Colex process--which uses a mercury solvent and lithium hydroxide as the lithium feed material-was shut down in 1962 and drained of process materials. Residual Quantities of mercury and lithium hydroxide have remained in the process equipment. Alpha-4 contains more than one-half million ft{sup 2} of floor area; 15,000 tons of process and electrical equipment; and 23,000 tons of insulation, mortar, brick, flooring, handrails, ducts, utilities, burnables, and sludge. Because much of this equipment and construction material is contaminated with elemental mercury, cleanup is necessary. The goal of the Y-12 Plant Decontamination and Decommissioning Technology Logic Diagram for Building 9201-4 is to provide a planning document that relates decontamination and decommissioning and waste management problems at the Alpha-4 building to the technologies that can be used to remediate these problems. The Y-12 Plant Decontamination and Decommissioning Technology Logic Diagram for Building 9201-4 builds on the methodology transferred by the U.S. Air Force to the Environmental Management organization with DOE and draws from previous technology logic diagram-efforts: logic diagrams for Hanford, the K-25 Site, and ORNL.

NONE

1994-09-01T23:59:59.000Z

295

2011 Uranium Marketing Annual Report - U.S. Energy Information ...  

U.S. Energy Information Administration (EIA)

Uranium Feed, Enrichment Services, Uranium Loaded In 2011, COOs delivered 51 million pounds U 3 O 8 e of natural uranium feed to U.S. and foreign enrichers. Fifty-

296

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

297

Safeguards Guidance Document for Designers of Commercial Nuclear Facilities: International Nuclear Safeguards Requirements and Practices For Uranium Enrichment Plants  

Science Conference Proceedings (OSTI)

This report is the second in a series of guidelines on international safeguards requirements and practices, prepared expressly for the designers of nuclear facilities. The first document in this series is the description of generic international nuclear safeguards requirements pertaining to all types of facilities. These requirements should be understood and considered at the earliest stages of facility design as part of a new process called Safeguards-by-Design. This will help eliminate the costly retrofit of facilities that has occurred in the past to accommodate nuclear safeguards verification activities. The following summarizes the requirements for international nuclear safeguards implementation at enrichment plants, prepared under the Safeguards by Design project, and funded by the U.S. Department of Energy (DOE) National Nuclear Security Administration (NNSA), Office of NA-243. The purpose of this is to provide designers of nuclear facilities around the world with a simplified set of design requirements and the most common practices for meeting them. The foundation for these requirements is the international safeguards agreement between the country and the International Atomic Energy Agency (IAEA), pursuant to the Treaty on the Non-proliferation of Nuclear Weapons (NPT). Relevant safeguards requirements are also cited from the Safeguards Criteria for inspecting enrichment plants, found in the IAEA Safeguards Manual, Part SMC-8. IAEA definitions and terms are based on the IAEA Safeguards Glossary, published in 2002. The most current specification for safeguards measurement accuracy is found in the IAEA document STR-327, International Target Values 2000 for Measurement Uncertainties in Safeguarding Nuclear Materials, published in 2001. For this guide to be easier for the designer to use, the requirements have been restated in plainer language per expert interpretation using the source documents noted. The safeguards agreement is fundamentally a legal document. As such, it is written in a legalese that is understood by specialists in international law and treaties, but not by most outside of this field, including designers of nuclear facilities. For this reason, many of the requirements have been simplified and restated. However, in all cases, the relevant source document and passage is noted so that readers may trace the requirement to the source. This is a helpful living guide, since some of these requirements are subject to revision over time. More importantly, the practices by which the requirements are met are continuously modernized by the IAEA and nuclear facility operators to improve not only the effectiveness of international nuclear safeguards, but also the efficiency. As these improvements are made, the following guidelines should be updated and revised accordingly.

Robert Bean; Casey Durst

2009-10-01T23:59:59.000Z

298

Verification experiment on the downblending of high enriched uranium (HEU) at the Portsmouth Gaseous Diffusion Plant. Digital video surveillance of the HEU feed stations  

SciTech Connect

As part of a Safeguards Agreement between the US and the International Atomic Energy Agency (IAEA), the Portsmouth Gaseous Diffusion Plant, Piketon, Ohio, was added to the list of facilities eligible for the application of IAEA safeguards. Currently, the facility is in the process of downblending excess inventory of HEU to low enriched uranium (LEU) from US defense related programs for commercial use. An agreement was reached between the US and the IAEA that would allow the IAEA to conduct an independent verification experiment at the Portsmouth facility, resulting in the confirmation that the HEU was in fact downblended. The experiment provided an opportunity for the DOE laboratories to recommend solutions/measures for new IAEA safeguards applications. One of the measures recommended by Sandia National Laboratories (SNL), and selected by the IAEA, was a digital video surveillance system for monitoring activity at the HEU feed stations. This paper describes the SNL implementation of the digital video system and its integration with the Load Cell Based Weighing System (LCBWS) from Oak Ridge National Laboratory (ORNL). The implementation was based on commercially available technology that also satisfied IAEA criteria for tamper protection and data authentication. The core of the Portsmouth digital video surveillance system was based on two Digital Camera Modules (DMC-14) from Neumann Consultants, Germany.

Martinez, R.L.; Tolk, K. [Sandia National Labs., Albuquerque, NM (United States); Whiting, N. [International Atomic Energy Agency, Vienna (Austria); Castleberry, K.; Lenarduzzi, R. [Oak Ridge National Lab., TN (United States)

1998-08-01T23:59:59.000Z

299

Uranium Industry Annual, 1992  

Science Conference Proceedings (OSTI)

The Uranium Industry Annual provides current statistical data on the US uranium industry for the Congress, Federal and State agencies, the uranium and electric utility industries, and the public. The feature article, ``Decommissioning of US Conventional Uranium Production Centers,`` is included. Data on uranium raw materials activities including exploration activities and expenditures, resources and reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities including domestic uranium purchases, commitments by utilities, procurement arrangements, uranium imports under purchase contracts and exports, deliveries to enrichment suppliers, inventories, secondary market activities, utility market requirements, and uranium for sale by domestic suppliers are presented in Chapter 2.

Not Available

1993-10-28T23:59:59.000Z

300

OXYGEN DIFFUSION IN HYPOSTOICHIOMETRIC URANIUM DIOXIDE  

E-Print Network (OSTI)

IN HYPOSTOICHIOMETRIC URANIUM DIOXIDE Kee Chul Kim Ph.D.727-366; Figure 1. Oxygen-uranium phase-equilibrium _ystem [18]. uranium dioxide powders and 18 0 enriched carbon

Kim, Kee Chul

2010-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "uranium enrichment decontamination" 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

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 ...

302

Uranium tailings bibliography  

SciTech Connect

A bibliography containing 1,212 references is presented with its focus on the general problem of reducing human exposure to the radionuclides contained in the tailings from the milling of uranium ore. The references are divided into seven broad categories: uranium tailings pile (problems and perspectives), standards and philosophy, etiology of radiation effects, internal dosimetry and metabolism, environmental transport, background sources of tailings radionuclides, and large-area decontamination. (JSR)

Holoway, C.F.; Goldsmith, W.A.; Eldridge, V.M.

1975-12-01T23:59:59.000Z

303

Preconceptual design of the gas-phase decontamination demonstration cart  

Science Conference Proceedings (OSTI)

Removal of uranium deposits from the interior surfaces of gaseous diffusion equipment will be a major portion of the overall multibillion dollar effort to decontaminate and decommission the gaseous diffusion plants. Long-term low-temperature (LTLT) gas-phase decontamination is being developed at the K-25 Site as an in situ decontamination process that is expected to significantly lower the decontamination costs, reduce worker exposure to radioactive materials, and reduce safeguard concerns. This report documents the preconceptual design of the process equipment that is necessary to conduct a full-scale demonstration of the LTLT method in accordance with the process steps listed above. The process equipment and method proposed in this report are not intended to represent a full-scale production campaign design and operation, since the gas evacuation, gas charging, and off-gas handling systems that would be cost effective in a production campaign are not cost effective for a first-time demonstration. However, the design presented here is expected to be applicable to special decontamination projects beyond the demonstration, which could include the Deposit Recovery Program. The equipment will therefore be sized to a 200 ft size 1 converter (plus a substantial conservative design margin), which is the largest item of interest for gas phase decontamination in the Deposit Recovery Program. The decontamination equipment will allow recovery of the UF{sub 6}, which is generated from the reaction of ClF{sub 3} with the uranium deposits, by use of NaF traps.

Munday, E.B.

1993-12-01T23:59:59.000Z

304

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

305

Uranium industry annual 1994  

SciTech Connect

The Uranium Industry Annual 1994 (UIA 1994) provides current statistical data on the US uranium industry`s activities relating to uranium raw materials and uranium marketing during that survey year. The UIA 1994 is prepared for use by the Congress, Federal and State agencies, the uranium and nuclear electric utility industries, and the public. It contains data for the 10-year period 1985 through 1994 as collected on the Form EIA-858, ``Uranium Industry Annual Survey.`` Data collected on the ``Uranium Industry Annual Survey`` (UIAS) provide a comprehensive statistical characterization of the industry`s activities for the survey year and also include some information about industry`s plans and commitments for the near-term future. Where aggregate data are presented in the UIA 1994, care has been taken to protect the confidentiality of company-specific information while still conveying accurate and complete statistical data. A feature article, ``Comparison of Uranium Mill Tailings Reclamation in the United States and Canada,`` is included in the UIA 1994. Data on uranium raw materials activities including exploration activities and expenditures, EIA-estimated resources and reserves, mine production of uranium, production of uranium concentrate, and industry employment are presented in Chapter 1. Data on uranium marketing activities, including purchases of uranium and enrichment services, and uranium inventories, enrichment feed deliveries (actual and projected), and unfilled market requirements are shown in Chapter 2.

NONE

1995-07-05T23:59:59.000Z

306

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

307

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.

308

Decontamination and reuse of ORGDP aluminum scrap  

Science Conference Proceedings (OSTI)

The Gaseous Diffusion Plants, or GDPs, have significant amounts of a number of metals, including nickel, aluminum, copper, and steel. Aluminum was used extensively throughout the GDPs because of its excellent strength to weight ratios and good resistance to corrosion by UF{sub 6}. This report is concerned with the recycle of aluminum stator and rotor blades from axial compressors. Most of the stator and rotor blades were made from 214-X aluminum casting alloy. Used compressor blades were contaminated with uranium both as a result of surface contamination and as an accumulation held in surface-connected voids inside of the blades. A variety of GDP studies were performed to evaluate the amounts of uranium retained in the blades; the volume, area, and location of voids in the blades; and connections between surface defects and voids. Based on experimental data on deposition, uranium content of the blades is 0.3%, or roughly 200 times the value expected from blade surface area. However, this value does correlate with estimated internal surface area and with lengthy deposition times. Based on a literature search, it appears that gaseous decontamination or melt refining using fluxes specific for uranium removal have the potential for removing internal contamination from aluminum blades. A melt refining process was used to recycle blades during the 1950s and 1960s. The process removed roughly one-third of the uranium from the blades. Blade cast from recycled aluminum appeared to perform as well as blades from virgin material. New melt refining and gaseous decontamination processes have been shown to provide substantially better decontamination of pure aluminum. If these techniques can be successfully adapted to treat aluminum 214-X alloy, internal and, possibly, external reuse of aluminum alloys may be possible.

Compere, A.L.; Griffith, W.L.; Hayden, H.W.; Wilson, D.F.

1996-12-01T23:59:59.000Z

309

Decontaminating Flooded Wells  

E-Print Network (OSTI)

This publication explains how to decontaminate and disinfect a well, test the well water and check for well damage after a flood.

Boellstorff, Diana; Dozier, Monty; Provin, Tony; Dictson, Nikkoal; McFarland, Mark L.

2005-09-30T23:59:59.000Z

310

Uranium Hexafluoride (UF6)  

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

Hexafluoride (UF6) Hexafluoride (UF6) Uranium Hexafluoride (UF6) line line Properties of UF6 UF6 Health Effects Uranium Hexafluoride (UF6) Physical and chemical properties of UF6, and its use in uranium processing. Uranium Hexafluoride and Its Properties Uranium hexafluoride is a chemical compound consisting of one atom of uranium combined with six atoms of fluorine. It is the chemical form of uranium that is used during the uranium enrichment process. Within a reasonable range of temperature and pressure, it can be a solid, liquid, or gas. Solid UF6 is a white, dense, crystalline material that resembles rock salt. UF6 crystals in a glass vial image UF6 crystals in a glass vial. Uranium hexafluoride does not react with oxygen, nitrogen, carbon dioxide, or dry air, but it does react with water or water vapor. For this reason,

311

Electrodic voltages accompanying stimulated bioremediation of a uranium-contaminated aquifer  

E-Print Network (OSTI)

Enrichment of members of the family Geobacteraceae associated with stimulation of dissimilatory metal reduction in uranium-

Williams, K.H.

2010-01-01T23:59:59.000Z

312

Bacterial Community Succession During in situ Uranium Bioremediation: Spatial Similarities Along Controlled Flow Paths  

E-Print Network (OSTI)

Enrichment of members of the family Geobacteraceae associated with stimulation of dissimilatory metal reduction in uranium-

Hwang, Chiachi

2009-01-01T23:59:59.000Z

313

Effects of organic carbon supply rates on mobility of previously bioreduced uranium in a contaminated sediment  

E-Print Network (OSTI)

Enrichment of members of the family Geobacteraceae associated with stimulation of dissimilatory metal reduction in uranium-

Wan, J.

2008-01-01T23:59:59.000Z

314

In-well sediment incubators to evaluate microbial community stability and dynamics following bioimmobilization of uranium  

E-Print Network (OSTI)

Enrichment of members of the family Geobacteraceae associated with stimulation of dissimilatory metal reduction in uranium-

Baldwin, B.R.

2010-01-01T23:59:59.000Z

315

Molecular analysis of phosphate limitation in Geobacteraceae during the bioremediation of a uranium-contaminated aquifer  

E-Print Network (OSTI)

Enrichment of members of the family Geobacteraceae associated with stimulation of dissimilatory metal reduction in uranium-

N'Guessan, L.A.

2010-01-01T23:59:59.000Z

316

Identifying the sources of subsurface contamination at the Hanford site in Washington using high-precision uranium isotopic measurements  

E-Print Network (OSTI)

enrichment for nuclear applications or to changes resulting from the burn up of natural or enriched uranium

Christensen, John N.; Dresel, P. Evan; Conrad, Mark E.; Maher, Kate; DePaolo, Donald J.

2004-01-01T23:59:59.000Z

317

FURTHER CONTINUING APPROPRIATIONS AMENDMENTS ...  

Science Conference Proceedings (OSTI)

... Atomic Energy Defense Activities National Nuclear Security Administration ... to the 'Uranium Enrichment Decontamination and Decommissioning ...

2012-01-24T23:59:59.000Z

319

Depleted uranium valuation  

SciTech Connect

The following uses for depleted uranium were examined to determine its value: a substitute for lead in shielding applications, feed material in gaseous diffusion enrichment facilities, feed material for an advanced enrichment concept, Mixed Oxide (MOx) diluent and blanket material in LMFBRs, and fertile material in LMFBR systems. A range of depleted uranium values was calculated for each of these applications. The sensitivity of these values to analysis assumptions is discussed. 9 tables.

Lewallen, M.A.; White, M.K.; Jenquin, U.P.

1979-04-01T23:59:59.000Z

320

Uranium purchases report 1994  

SciTech Connect

US utilities are required to report to the Secretary of Energy annually the country of origin and the seller of any uranium or enriched uranium purchased or imported into the US, as well as the country of origin and seller of any enrichment services purchased by the utility. This report compiles these data and also contains a glossary of terms and additional purchase information covering average price and contract duration. 3 tabs.

1995-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "uranium enrichment decontamination" 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.


321

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

322

Long lasting decontamination foam  

DOE Patents (OSTI)

Compositions and methods for decontaminating surfaces are disclosed. More specifically, compositions and methods for decontamination using a composition capable of generating a long lasting foam are disclosed. Compositions may include a surfactant and gelatin and have a pH of less than about 6. Such compositions may further include affinity-shifting chemicals. Methods may include decontaminating a contaminated surface with a composition or a foam that may include a surfactant and gelatin and have a pH of less than about 6.

Demmer, Ricky L. (Idaho Falls, ID); Peterman, Dean R. (Idaho Falls, ID); Tripp, Julia L. (Pocatello, ID); Cooper, David C. (Idaho Falls, ID); Wright, Karen E. (Idaho Falls, ID)

2010-12-07T23:59:59.000Z

323

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...

324

FAQ 9-Where does uranium hexafluoride come from?  

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

hexafluoride come from? Where does uranium hexafluoride come from? The gaseous diffusion process used to enrich uranium requires uranium in the form of UF6. In the first step of...

325

Contaminant-Organic Complexes: Their Structure and Energetics in Surface Decontamination Processes  

SciTech Connect

The current debate over possible decontamination processes for U.S. Department of Energy (DOE) facilities is centered on disparate decontamination problems, but the key contaminants (uranium [U], plutonium [Pu], and neptunium [Np]) are universally important. There is no single decontamination technique or agent for all metal surfaces and contaminants with which DOE is faced. However, more innovative agents used alone or in conjunction with traditional processes can increase the potential to reclaim for future use some of these valuable resources or, at the least, decontaminate the metal surfaces to allow disposal as nonradioactive, nonhazardous material. This debate underscores several important issues: (1) regardless of the decontamination scenario, metal (Fe, U, Pu, Np) oxide film removal from the surface is central to decontamination; and (2) simultaneous oxide dissolution and sequestration of actinide contaminants against re-adsorption to a clean metal surface will influence the efficacy of a process or agent and its cost.

Ainsworth, Calvin C.; Hay, Benjamin P.; Traina, Samuel J.; Myneni, Satish C. B.

2002-06-01T23:59:59.000Z

326

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

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

Flickr RSS Twitter YouTube GTRI's Convert Program: Minimizing the Use of Highly Enriched Uranium | National Nuclear Security Administration Our Mission Managing the Stockpile...

327

NNSA and Kazakhstan Complete Operation to Eliminate Highly Enriched...  

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

Flickr RSS Twitter YouTube NNSA and Kazakhstan Complete Operation to Eliminate Highly Enriched Uranium | National Nuclear Security Administration Our Mission Managing the Stockpile...

328

METHOD AND APPARATUS FOR MEASURING ENRICHMENT OF UF6 - Energy ...  

A system and method are disclosed for determining the enrichment of .sup.235U in Uranium Hexafluoride (UF6) utilizing synthesized X-rays which are ...

329

Oxidative Tritium Decontamination System  

DOE Patents (OSTI)

The Oxidative Tritium Decontamination System, OTDS, provides a method and apparatus for reduction of tritium surface contamination on various items. The OTDS employs ozone gas as oxidizing agent to convert elemental tritium to tritium oxide. Tritium oxide vapor and excess ozone gas is purged from the OTDS, for discharge to atmosphere or transport to further process. An effluent stream is subjected to a catalytic process for the decomposition of excess ozone to diatomic oxygen. One of two configurations of the OTDS is employed: dynamic apparatus equipped with agitation mechanism and large volumetric capacity for decontamination of light items, or static apparatus equipped with pressurization and evacuation capability for decontamination of heavier, delicate, and/or valuable items.

Gentile, Charles A. (Plainsboro, NJ), Guttadora, Gregory L. (Highland Park, NJ), Parker, John J. (Medford, NJ)

2006-02-07T23:59:59.000Z

330

Sulfur isotopes as indicators of amended bacterial sulfate reduction processes influencing field scale uranium bioremediation  

E-Print Network (OSTI)

sulfate and uranium bioreduction processes. Enrichment of upenrichment of 238 U relative to 235 U in residual U(VI) during microbial uranium

Druhan, J.L.

2009-01-01T23:59:59.000Z

331

Method And Apparatus For Measuring Enrichment Of UF6  

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

For Measuring Enrichment Of UF6 A system and method are disclosed for determining the enrichment of .sup.235U in Uranium Hexafluoride (UF6) utilizing synthesized X-rays which...

332

FAQ 7-How is depleted uranium produced?  

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

How is depleted uranium produced? How is depleted uranium produced? How is depleted uranium produced? Depleted uranium is produced during the uranium enrichment process. In the United States, uranium is enriched through the gaseous diffusion process in which the compound uranium hexafluoride (UF6) is heated and converted from a solid to a gas. The gas is then forced through a series of compressors and converters that contain porous barriers. Because uranium-235 has a slightly lighter isotopic mass than uranium-238, UF6 molecules made with uranium-235 diffuse through the barriers at a slightly higher rate than the molecules containing uranium-238. At the end of the process, there are two UF6 streams, with one stream having a higher concentration of uranium-235 than the other. The stream having the greater uranium-235 concentration is referred to as enriched UF6, while the stream that is reduced in its concentration of uranium-235 is referred to as depleted UF6. The depleted UF6 can be converted to other chemical forms, such as depleted uranium oxide or depleted uranium metal.

333

RUTHENIUM DECONTAMINATION METHOD  

DOE Patents (OSTI)

A liquid-liquid extraction method of separating uranium from fission products is given. A small amount of a low molecular weight ketone is added to an acidic aqueous solution containing neutron-irradiated uranium and its associated fission products. The resulting solution is digested and then contacted with an organic liquid that extracts uranium values. The purpose of the step of digesting the aqueous solution in the presence of the ketone is to suppress the extractability of ruthenium.

Gresky, A.T.

1960-07-19T23:59:59.000Z

334

Decontaminating metal surfaces  

DOE Patents (OSTI)

Radioactively contaminated surfaces can be electrolytically decontaminated with greatly increased efficiencies by using electrolytes containing higher than heretofore conventional amounts of nitrate, e.g.,>600 g/l of NaNO.sub.3, or by using nitrate-containing electrolytes which are acidic, e.g., of a pH<6.

Childs, Everett L. (Boulder, CO)

1984-11-06T23:59:59.000Z

335

Decontaminating metal surfaces  

DOE Patents (OSTI)

Radioactively contaminated surfaces can be electrolytically decontaminated with greatly increased efficiencies by using electrolytes containing higher than heretofore conventional amounts of nitrate, e.g., >600 g/1 of NaNO/sub 3/, or by using nitrate-containing electrolytes which are acidic, e.g., of a pH < 6.

Childs, E.L.

1984-01-23T23:59:59.000Z

336

Electrolytic decontamination of conductive materials for hazardous waste management  

SciTech Connect

Electrolytic removal of plutonium and americium from stainless steel and uranium surfaces has been demonstrated. Preliminary experiments were performed on the electrochemically based decontamination of type 304L stainless steel in sodium nitrate solutions to better understand the metal removal effects of varying cur-rent density, pH, and nitrate concentration parameters. Material removal rates and changes in surface morphology under these varying conditions are reported. Experimental results indicate that an electropolishing step before contamination removes surface roughness, thereby simplifying later electrolytic decontamination. Sodium nitrate based electrolytic decontamination produced the most uniform stripping of material at low to intermediate pH and at sodium nitrate concentrations of 200 g L{sup -1} and higher. Stirring was also observed to increase the uniformity of the stripping process.

Wedman, D.E.; Martinez, H.E.; Nelson, T.O.

1996-12-31T23:59:59.000Z

337

Beneficial Uses of Depleted Uranium  

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

Table 2 (ref. 1). The content of 235 U in DU is dependent on economics. If the cost of natural uranium feed is high relative to the cost of enrichment services, then a low 235 U...

338

Radiological decontamination, survey, and statistical release method for vehicles  

SciTech Connect

Earth-moving vehicles (e.g., dump trucks, belly dumps) commonly haul radiologically contaminated materials from a site being remediated to a disposal site. Traditionally, each vehicle must be surveyed before being released. The logistical difficulties of implementing the traditional approach on a large scale demand that an alternative be devised. A statistical method for assessing product quality from a continuous process was adapted to the vehicle decontamination process. This method produced a sampling scheme that automatically compensates and accommodates fluctuating batch sizes and changing conditions without the need to modify or rectify the sampling scheme in the field. Vehicles are randomly selected (sampled) upon completion of the decontamination process to be surveyed for residual radioactive surface contamination. The frequency of sampling is based on the expected number of vehicles passing through the decontamination process in a given period and the confidence level desired. This process has been successfully used for 1 year at the former uranium millsite in Monticello, Utah (a cleanup site regulated under the Comprehensive Environmental Response, Compensation, and Liability Act). The method forces improvement in the quality of the decontamination process and results in a lower likelihood that vehicles exceeding the surface contamination standards are offered for survey. Implementation of this statistical sampling method on Monticello projects has resulted in more efficient processing of vehicles through decontamination and radiological release, saved hundreds of hours of processing time, provided a high level of confidence that release limits are met, and improved the radiological cleanliness of vehicles leaving the controlled site.

Goodwill, M.E.; Lively, J.W.; Morris, R.L.

1996-06-01T23:59:59.000Z

339

DOE Announces Transfer of Depleted Uranium to Advance the U.S...  

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

low-enriched uranium related to NNSA's programs for down-blending surplus U.S. highly enriched uranium. Based on this analysis, Secretary Chu made a determination that the above...

340

DETECTION OF ULTRA-TRACE LEVELS OF URANIUM IN AQUEOUS SAMPLES BY LASER INDUCED FLUORESCENCE SPECTROMETRY  

E-Print Network (OSTI)

uranium have been reported in the chemical literature (1), many of which involve initial separation or enrichment

Perry, Dale L.

2012-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "uranium enrichment decontamination" 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

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

342

DECONTAMINATION AND DECOMMISSIONING AT THEEAST TENNESSEE TECHNOLOGYPAR...  

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

DECONTAMINATION AND DECOMMISSIONING AT THEEAST TENNESSEE TECHNOLOGYPARK, ER-B-99-01 DECONTAMINATION AND DECOMMISSIONING AT THEEAST TENNESSEE TECHNOLOGYPARK, ER-B-99-01 The East...

343

Toxic Chemical Agent Decontamination Emulsions, Their ...  

U.S. Energy Information Administration (EIA)

This invention is related to decontaminating agents and amethod for the decontamination of ... which have been contaminated with toxic chemical agents ...

344

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 ...

345

Decontamination solution development studies  

SciTech Connect

This study was conducted for the Westinghouse Hanford Company (WHC) by Pacific Northwest Laboratory (PNL) as part of the Hanford Grout Technology Program (HGTP). The objective of this study was to identify decontamination solutions capable of removing radioactive contaminants and grout from the Grout Treatment Facility (GTF) process equipment and to determine the impact of these solutions on equipment components and disposal options. The reference grout used in this study was prepared with simulated double-shell slurry feed (DSSF) and a dry blend consisting of 40 wt % limestone flour, 28 wt % blast furnace slag, 28 wt % fly ash, and 4 wt % type I/II Portland cement.

Allen, R.P.; Fetrow, L.K.; Kjarmo, H.E.; Pool, K.H.

1993-09-01T23:59:59.000Z

346

Integrated decontamination process for metals  

DOE Patents (OSTI)

An integrated process for decontamination of metals, particularly metals that are used in the nuclear energy industry contaminated with radioactive material. The process combines the processes of electrorefining and melt refining to purify metals that can be decontaminated using either electrorefining or melt refining processes.

Snyder, Thomas S. (Oakmont, PA); Whitlow, Graham A. (Murrysville, PA)

1991-01-01T23:59:59.000Z

347

DOE Announces Policy for Managing Excess Uranium Inventory | Department of  

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

Policy for Managing Excess Uranium Inventory Policy for Managing Excess Uranium Inventory DOE Announces Policy for Managing Excess Uranium Inventory March 12, 2008 - 10:52am Addthis WASHINGTON, DC - U.S. Secretary of Energy Samuel W. Bodman today released a Policy Statement on the management of the Department of Energy's (DOE) excess uranium inventory, providing the framework within which DOE will make decisions concerning future use and disposition of its inventory. During the coming year, DOE will continue its ongoing program for downblending excess highly enriched uranium (HEU) into low enriched uranium (LEU), evaluate the benefits of enriching a portion of its excess natural uranium into LEU, and complete an analysis on enriching and/or selling some of its depleted uranium. Specific transactions are expected to occur in

348

Uranium at Y-12: Inspection | Y-12 National Security Complex  

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

Inspection Uranium at Y-12: Inspection Posted: July 22, 2013 - 3:36pm | Y-12 Report | Volume 10, Issue 1 | 2013 Inspection of enriched uranium is performed by dimensional checks...

349

Feasibility of gas-phase decontamination of gaseous diffusion equipment  

SciTech Connect

The five buildings at the K-25 Site formerly involved in the gaseous diffusion process contain 5000 gaseous diffusion stages as well as support facilities that are internally contaminated with uranium deposits. The gaseous diffusion facilities located at the Portsmouth Gaseous Diffusion Plant and the Paducah Gaseous Diffusion Plant also contain similar equipment and will eventually close. The decontamination of these facilities will require the most cost-effective technology consistent with the criticality, health physics, industrial hygiene, and environmental concerns; the technology must keep exposures to hazardous substances to levels as low as reasonably achievable (ALARA). This report documents recent laboratory experiments that were conducted to determine the feasibility of gas-phase decontamination of the internal surfaces of the gaseous diffusion equipment that is contaminated with uranium deposits. A gaseous fluorinating agent is used to fluorinate the solid uranium deposits to gaseous uranium hexafluoride (UF{sub 6}), which can be recovered by chemical trapping or freezing. The lab results regarding the feasibility of the gas-phase process are encouraging. These results especially showed promise for a novel decontamination approach called the long-term, low-temperature (LTLT) process. In the LTLT process: The equipment is rendered leak tight, evacuated, leak tested, and pretreated, charged with chlorine trifluoride (ClF{sub 3}) to subatmospheric pressure, left for an extended period, possibly > 4 months, while processing other items. Then the UF{sub 6} and other gases are evacuated. The UF{sub 6} is recovered by chemical trapping. The lab results demonstrated that ClF{sub 3} gas at subatmospheric pressure and at {approx} 75{degree}F is capable of volatilizing heavy deposits of uranyl fluoride from copper metal surfaces sufficiently that the remaining radioactive emissions are below limits.

Munday, E.B.; Simmons, D.W.

1993-02-01T23:59:59.000Z

350

Calibration of the Lawrence Livermore National Laboratory Passive-Active Neutron Drum Shuffler for Measurement of Highly Enriched Uranium in Oxides within DOE-STD-3013-2000 Containers  

SciTech Connect

Lawrence Livermore National Laboratory (LLNL) uses 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. In June 2002, at the 43rd Annual Meeting of the Institute of Nuclear Material Management, LLNL reported on an extensive effort to calibrate this shuffler, based on standards measurements and extensive simulations, for HEU oxides and mixed U-Pu oxides in thin-walled primary and secondary containers. In August 2002, LLNL began to also use DOE-STD-3013-2000 containers for HEU oxide and mixed U-Pu oxide. These DOE-STD-3013-2000 containers are comprised of a stainless steel convenience can enclosed in welded stainless steel primary and secondary containers. Compared to the double thin-walled containers, the DOE-STD-3013-2000 containers have substantially thicker walls, and the density of materials in these containers was found to extend over a greater range (1.35 g/cm{sup 3} to 4.62 g/cm{sup 3}) than foreseen for the double thin-walled containers. Further, the DOE-STD-3013-2000 Standard allows for oxides containing at least 30 wt% Pu plus U whereas the calibration algorithms for thin-walled containers were derived for virtually pure HEU or mixed U-Pu oxides. An initial series of Monte Carlo simulations of the PAN shuffler response to given quantities of HEU oxide and mixed U-Pu oxide in DOE-STD-3013-2000 containers was generated and compared with the response predicted by the calibration algorithms for thin-walled containers. Results showed a decrease on the order of 10% in the count rate, and hence a decrease in the calculated U mass for measured unknowns, with some varying trends versus U mass. Therefore a decision was made to develop a calibration algorithm for the PAN shuffler unique to the DOE-STD-3013-2000 container. This paper describes that effort and selected unknown item measurement results.

Mount, M E; O' Connell, W J

2005-06-03T23:59:59.000Z

351

An Evaluation of Uranium Measurement Capabilities and Comparison  

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

many analytical techniques and actinide material forms. NBL also does work in highly enriched uranium transparency monitoring, assists in Material Control and Accountability...

352

Vector Representation as a Tool for Detecting Characteristic Uranium Peaks.  

E-Print Network (OSTI)

??Vector representation is found as a viable tool for identifying the presence of and determining the difference between enriched and naturally occurring uranium. This was (more)

Forney, Anne Marie

2012-01-01T23:59:59.000Z

353

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 ...

354

Chemical agent decontamination composition comprising a ...  

U.S. Energy Information Administration (EIA)

Title: Chemical agent decontamination composition comprising a perfluorinated alkyl bromide Date: 05/13/2008

355

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

356

Glovebox decontamination technology comparison  

SciTech Connect

Reconfiguration of the CMR Building and TA-55 Plutonium Facility for mission requirements will require the disposal or recycle of 200--300 gloveboxes or open front hoods. These gloveboxes and open front hoods must be decontaminated to meet discharge limits for Low Level Waste. Gloveboxes and open front hoods at CMR have been painted. One of the deliverables on this project is to identify the best method for stripping the paint from large numbers of gloveboxes. Four methods being considered are the following: conventional paint stripping, dry ice pellets, strippable coatings, and high pressure water technology. The advantages of each technology will be discussed. Last, cost comparisons between the technologies will be presented.

Quintana, D.M.; Rodriguez, J.B.; Cournoyer, M.E.

1999-09-26T23:59:59.000Z

357

Decontamination & decommissioning focus area  

Science Conference Proceedings (OSTI)

In January 1994, the US Department of Energy Office of Environmental Management (DOE EM) formally introduced its new approach to managing DOE`s environmental research and technology development activities. The goal of the new approach is to conduct research and development in critical areas of interest to DOE, utilizing the best talent in the Department and in the national science community. To facilitate this solutions-oriented approach, the Office of Science and Technology (EM-50, formerly the Office of Technology Development) formed five Focus AReas to stimulate the required basic research, development, and demonstration efforts to seek new, innovative cleanup methods. In February 1995, EM-50 selected the DOE Morgantown Energy Technology Center (METC) to lead implementation of one of these Focus Areas: the Decontamination and Decommissioning (D & D) Focus Area.

NONE

1996-08-01T23:59:59.000Z

358

New generation enrichment monitoring technology for gas centrifuge enrichment plants  

SciTech Connect

The continuous enrichment monitor, developed and fielded in the 1990s by the International Atomic Energy Agency, provided a go-no-go capability to distinguish between UF{sub 6} containing low enriched (approximately 4% {sup 235}U) and highly enriched (above 20% {sup 235}U) uranium. This instrument used the 22-keV line from a {sup 109}Cd source as a transmission source to achieve a high sensitivity to the UF{sub 6} gas absorption. The 1.27-yr half-life required that the source be periodically replaced and the instrument recalibrated. The instrument's functionality and accuracy were limited by the fact that measured gas density and gas pressure were treated as confidential facility information. The modern safeguarding of a gas centrifuge enrichment plant producing low-enriched UF{sub 6} product aims toward a more quantitative flow and enrichment monitoring concept that sets new standards for accuracy stability, and confidence. An instrument must be accurate enough to detect the diversion of a significant quantity of material, have virtually zero false alarms, and protect the operator's proprietary process information. We discuss a new concept for advanced gas enrichment assay measurement technology. This design concept eliminates the need for the periodic replacement of a radioactive source as well as the need for maintenance by experts. Some initial experimental results will be presented.

Ianakiev, Kiril D [Los Alamos National Laboratory; Alexandrov, Boian, S. [Los Alamos National Laboratory; Boyer, Brian, D. [Los Alamos National Laboratory; Hill, Thomas, R. [Los Alamos National Laboratory; Macarthur, Duncan, W. [Los Alamos National Laboratory; Marks, Thomas [Los Alamos National Laboratory; Moss, Calvin, E. [Los Alamos National Laboratory; Sheppard, Gregory, A. [Los Alamos National Laboratory; Swinhoe, Martyn, T. [Los Alamos National Laboratory

2008-01-01T23:59:59.000Z

359

Commercializationof Dredged-Material Decontamination  

E-Print Network (OSTI)

loans for constructionoffacilitiesthat must run for a long-term period in order to amortize the capital costs. Considerationshouldbe givento the developmentof mechanismsthat could make long of sediment decontamination obtaining adequatefunding for capital and operating costs during the tecbnob

Brookhaven National Laboratory

360

DOE Signs Advanced Enrichment Technology License and Facility Lease |  

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

Advanced Enrichment Technology License and Facility Lease Advanced Enrichment Technology License and Facility Lease DOE Signs Advanced Enrichment Technology License and Facility Lease December 8, 2006 - 9:34am Addthis Announces Agreements with USEC Enabling Deployment of Advanced Domestic Technology for Uranium Enrichment WASHINGTON, DC - U.S. Secretary of Energy Samuel W. Bodman today announced the signing of a lease agreement with the United States Enrichment Corporation, Inc. (USEC) for their use of the Department's gas centrifuge enrichment plant (GCEP) facilities in Piketon, OH for their American Centrifuge Plant. The Department of Energy (DOE) also granted a non-exclusive patent license to USEC for use of DOE's centrifuge technology for uranium enrichment at the plant, which will initiate the first successful deployment of advanced domestic enrichment technology in the

Note: This page contains sample records for the topic "uranium enrichment decontamination" 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

From the lab to the real world : sources of error in UF6 gas enrichment monitoring.  

E-Print Network (OSTI)

??Safeguarding uranium enrichment facilities is a serious concern for the International Atomic Energy Agency (IAEA). Safeguards methods have changed over the years, most recently switching (more)

Lombardi, Marcie

2012-01-01T23:59:59.000Z

362

Withdrawal assay monitoring at US Enrichment Facilities  

SciTech Connect

The United States Enrichment Corporation (USEC) controls two uranium enrichment facilities that produce enriched uranium for both military and commercial use. The process requires both feed and withdrawal operations. The withdrawal process requires both product (enriched uranium) withdrawal stations and tails (depleted uranium) withdrawal stations. A previous prototype system, ``X-330 Tails Cylinder Assay Monitor,`` was developed as a demonstration for the tails withdrawal station at the Portsmouth Gaseous Diffusion Plant (PORTS). The prototype system was done in response to potential problems with the original method for determining the hourly weighted assay averages that are used to calculate the final weighted assay of the cylinder. In the original method the {sup 235}U assay of uranium hexaflouride withdrawn from PORTS cascade into tails cylinders is determined every 5 min by measurements from an in-line assay mass spectrometer. An average value for a 1-h period is then calculated by area control room personnel and assigned to the accumulated weight in the cylinder for the period. A potential problem with this method is that cylinder weight is not automatically recorded as often as the assay. The assay and withdrawal rate can both vary during the given period. This variation results in inaccuracies in the hourly weighted assays that are used to calculate the final weighted assay of the cylinder. Laboratory analysis is considered to be the most accurate method for determining the final cylinder assay; however, the cost and safety considerations of redundant cylinder handling limit the number of cylinders sampled to less than 10%.

Smith, D.E.

1996-01-01T23:59:59.000Z

363

India's Worsening Uranium Shortage  

Science Conference Proceedings (OSTI)

As a result of NSG restrictions, India cannot import the natural uranium required to fuel its Pressurized Heavy Water Reactors (PHWRs); consequently, it is forced to rely on the expediency of domestic uranium production. However, domestic production from mines and byproduct sources has not kept pace with demand from commercial reactors. This shortage has been officially confirmed by the Indian Planning Commissions Mid-Term Appraisal of the countrys current Five Year Plan. The report stresses that as a result of the uranium shortage, Indian PHWR load factors have been continually decreasing. The Uranium Corporation of India Ltd (UCIL) operates a number of underground mines in the Singhbhum Shear Zone of Jharkhand, and it is all processed at a single mill in Jaduguda. UCIL is attempting to aggrandize operations by establishing new mines and mills in other states, but the requisite permit-gathering and development time will defer production until at least 2009. A significant portion of Indias uranium comes from byproduct sources, but a number of these are derived from accumulated stores that are nearing exhaustion. A current maximum estimate of indigenous uranium production is 430t/yr (230t from mines and 200t from byproduct sources); whereas, the current uranium requirement for Indian PHWRs is 455t/yr (depending on plant capacity factor). This deficit is exacerbated by the additional requirements of the Indian weapons program. Present power generation capacity of Indian nuclear plants is 4350 MWe. The power generation target set by the Indian Department of Atomic Energy (DAE) is 20,000 MWe by the year 2020. It is expected that around half of this total will be provided by PHWRs using indigenously supplied uranium with the bulk of the remainder provided by breeder reactors or pressurized water reactors using imported low-enriched uranium.

Curtis, Michael M.

2007-01-15T23:59:59.000Z

364

Disposition of Surplus Highly Enriched Uranium  

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

. . ------- .--- --. ---- DOE/EIS-0240 I United States Department of Energy I For Further Information Contact: U.S. Department of Energy Otice of Fissile Materials Disposition, 1000 Independence Ave., SW, Washington, D.C. 20585 1 I ---- I I . I I I I This report has been reproduced directly from the best available copy. Available to DOE and DOE contractors from 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: I Office of Fissile Materials Disposition, MD-4 Forrestal Building United States Department of Energy 1000 Independence Avenue, SW Washington, DC 20585 , @ Printed with soy ink on recycled paper. -_. - COVERS~ET

365

Disposition of Surplus Highly Enriched Uranium  

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

EIS-0240-S EIS-0240-S For Further Information Contact: U.S. Departmel>t of Energy Office of Fissile Materials Disposition, 1000 Independence Ave., SW, Washington, D.C. 20585 . This report has been reproduced directly from the best available copy. Available to DOE and DOE contractors from 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 @ Printed with soy ink on recycled paper. .__- -. @ .: Depafimmt of Energy . i i~t " Wastin@on, DC 20585 June 1996 Dear hterested

366

SciTech Connect: enriched uranium  

Office of Scientific and Technical Information (OSTI)

Hydrogenation of Dislocation-Limited Heteroepitaxial Silicon Solar Cells: Preprint Citation Details In-Document Search Title: Hydrogenation of Dislocation-Limited Heteroepitaxial...

367

Highly Enriched Uranium Transparency Program | National Nuclear...  

National Nuclear Security Administration (NNSA)

Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure...

368

Highly Enriched Uranium Disposition | National Nuclear Security...  

National Nuclear Security Administration (NNSA)

the United States Senate Committee on Armed Services Sep 17, 2013 NNSA, Republic of Korea Ministry Agree to Minimize Use of HEU in Nuclear Reactors Sep 3, 2013 NNSA Conducts...

369

Uranium and Its Compounds  

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

and Its Compounds Uranium and Its Compounds line line What is Uranium? Chemical Forms of Uranium Properties of Uranium Compounds Radioactivity and Radiation Uranium Health Effects...

370

Isotopic ratio method for determining uranium contamination  

SciTech Connect

The presence of high concentrations of uranium in the subsurface can be attributed either to contamination from uranium processing activities or to naturally occurring uranium. A mathematical method has been employed to evaluate the isotope ratios from subsurface soils at the Rocky Flats Nuclear Weapons Plant (RFP) and demonstrates conclusively that the soil contains uranium from a natural source and has not been contaminated with enriched uranium resulting from RFP releases. This paper describes the method used in this determination which has widespread application in site characterizations and can be adapted to other radioisotopes used in manufacturing industries. The determination of radioisotope source can lead to a reduction of the remediation effort.

Miles, R.E.; Sieben, A.K.

1994-02-03T23:59:59.000Z

371

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

372

Inherently safe in situ uranium recovery.  

SciTech Connect

Expansion of uranium mining in the United States is a concern to some environmental groups and sovereign Native American Nations. An approach which may alleviate some problems is to develop inherently safe in situ uranium recovery ('ISR') technologies. Current ISR technology relies on chemical extraction of trace levels of uranium from aquifers that, once mined, can still contain dissolved uranium and other trace metals that are a health concern. Existing ISR operations are few in number; however, high uranium prices are driving the industry to consider expanding operations nation-wide. Environmental concerns and enforcement of the new 30 ppb uranium drinking water standard may make opening new mining operations more difficult and costly. Here we propose a technological fix: the development of inherently safe in situ recovery (ISISR) methods. The four central features of an ISISR approach are: (1) New 'green' leachants that break down predictably in the subsurface, leaving uranium, and associated trace metals, in an immobile form; (2) Post-leachant uranium/metals-immobilizing washes that provide a backup decontamination process; (3) An optimized well-field design that increases uranium recovery efficiency and minimizes excursions of contaminated water; and (4) A combined hydrologic/geochemical protocol for designing low-cost post-extraction long-term monitoring. ISISR would bring larger amounts of uranium to the surface, leave fewer toxic metals in the aquifer, and cost less to monitor safely - thus providing a 'win-win-win' solution to all stakeholders.

Krumhansl, James Lee; Beauheim, Richard Louis; Brady, Patrick Vane; Arnold, Bill Walter; Kanney, Joseph F.; McKenna, Sean Andrew

2009-05-01T23:59:59.000Z

373

Inherently safe in situ uranium recovery.  

SciTech Connect

Expansion of uranium mining in the United States is a concern to some environmental groups and sovereign Native American Nations. An approach which may alleviate some problems is to develop inherently safe in situ uranium recovery ('ISR') technologies. Current ISR technology relies on chemical extraction of trace levels of uranium from aquifers that, once mined, can still contain dissolved uranium and other trace metals that are a health concern. Existing ISR operations are few in number; however, high uranium prices are driving the industry to consider expanding operations nation-wide. Environmental concerns and enforcement of the new 30 ppb uranium drinking water standard may make opening new mining operations more difficult and costly. Here we propose a technological fix: the development of inherently safe in situ recovery (ISISR) methods. The four central features of an ISISR approach are: (1) New 'green' leachants that break down predictably in the subsurface, leaving uranium, and associated trace metals, in an immobile form; (2) Post-leachant uranium/metals-immobilizing washes that provide a backup decontamination process; (3) An optimized well-field design that increases uranium recovery efficiency and minimizes excursions of contaminated water; and (4) A combined hydrologic/geochemical protocol for designing low-cost post-extraction long-term monitoring. ISISR would bring larger amounts of uranium to the surface, leave fewer toxic metals in the aquifer, and cost less to monitor safely - thus providing a 'win-win-win' solution to all stakeholders.

Krumhansl, James Lee; Beauheim, Richard Louis; Brady, Patrick Vane; Arnold, Bill Walter; Kanney, Joseph F.; McKenna, Sean Andrew

2009-05-01T23:59:59.000Z

374

Development of the Decontamination Approach for the West Valley Demonstration Project Decontamination Project Plan  

SciTech Connect

This paper details the development of a decontamination approach for the West Valley Demonstration Project (WVDP), Decontamination Project Plan (Plan). The WVDP is operated by West Valley Nuclear Services Company (WVNSCO), a subsidiary of Westinghouse Government and Environmental Services, and its parent companies Washington Group International and British Nuclear Fuels Limited (BNFL). The WVDP is a waste management effort being conducted by the United States Department of Energy (DOE) at the site of the only commercial nuclear fuel reprocessing facility to have operated in the United States. This facility is part of the Western New York Nuclear Service Center (WNYNSC), which is owned by the New York State Energy Research and Development Authority (NYSERDA). As authorized by Congress in 1980 through the West Valley Demonstration Project Act (WVDP Act, Public Law 96-368), the DOE's primary mission at the WVDP is to solidify high-level liquid nuclear waste safely; transport the high-level waste (HLW) to a federal repository; and decontaminate and decommission the facilities and hardware used to solidify the HLW and conduct the WVDP. This includes a provision for the disposal of low-level waste (LLW) and transuranic waste (TRU) produced during processing of the HLW. Continuation of the effort to reduce the hazard and risk associated with historic operations to the extent needed to ensure the health and safety of the public and the environment will see a change in focus from stabilization of liquid HLW to stabilization of former plutonium and uranium extraction (PUREX) reprocessing plant facilities. This will be achieved through the activities of in-cell component removal and packaging, and preparation for long-term disposal of the long- lived radionuclides. These radionuclides are associated with the former PUREX facility operations, including, and upstream from, facilities utilized in the primary separation and first plutonium/uranium split cycles. The closure strategy for the WVDP is subject to ongoing evaluation and decision-making involving DOE and NYSERDA. Implementation will be subject to a future Record of Decision (ROD) and an Environmental Impact Statement (EIS).

Milner, T. N.; Watters, W. T.

2002-02-25T23:59:59.000Z

375

DECONTAMINATION TECHNOLOGIES FOR FACILITY REUSE  

SciTech Connect

As nuclear research and production facilities across the U.S. Department of Energy (DOE) nuclear weapons complex are slated for deactivation and decommissioning (D&D), there is a need to decontaminate some facilities for reuse for another mission or continued use for the same mission. Improved technologies available in the commercial sector and tested by the DOE can help solve the DOE's decontamination problems. Decontamination technologies include mechanical methods, such as shaving, scabbling, and blasting; application of chemicals; biological methods; and electrochemical techniques. Materials to be decontaminated are primarily concrete or metal. Concrete materials include walls, floors, ceilings, bio-shields, and fuel pools. Metallic materials include structural steel, valves, pipes, gloveboxes, reactors, and other equipment. Porous materials such as concrete can be contaminated throughout their structure, although contamination in concrete normally resides in the top quarter-inch below the surface. Metals are normally only contaminated on the surface. Contamination includes a variety of alpha, beta, and gamma-emitting radionuclides and can sometimes include heavy metals and organic contamination regulated by the Resource Conservation and Recovery Act (RCRA). This paper describes several advanced mechanical, chemical, and other methods to decontaminate structures, equipment, and materials.

Bossart, Steven J.; Blair, Danielle M.

2003-02-27T23:59:59.000Z

376

Uranium Metal: Potential for Discovering Commercial Uses  

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

Uranium Metal Uranium Metal Potential for Discovering Commercial Uses Steven M. Baker, Ph.D. Knoxville Tn 5 August 1998 Summary Uranium Metal is a Valuable Resource 3 Large Inventory of "Depleted Uranium" 3 Need Commercial Uses for Inventory  Avoid Disposal Cost  Real Added Value to Society 3 Uranium Metal Has Valuable Properties  Density  Strength 3 Market will Come if Story is Told Background The Nature of Uranium Background 3 Natural Uranium: 99.3% U238; 0.7% U 235 3 U235 Fissile  Nuclear Weapons  Nuclear Reactors 3 U238 Fertile  Neutron Irradiation of U238 Produces Pu239  Neutrons Come From U235 Fission  Pu239 is Fissile (Weapons, Reactors, etc.) Post World War II Legacy Background 3 "Enriched" Uranium Product  Weapons Program 

377

Paducah Plant Begins Enrichment Operations after Five Parties Strike  

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

Plant Begins Enrichment Operations after Five Parties Plant Begins Enrichment Operations after Five Parties Strike Agreement Paducah Plant Begins Enrichment Operations after Five Parties Strike Agreement May 1, 2012 - 12:00pm Addthis This cylinder hauler at Paducah’s Babcock & Wilcox Conversion Services plant delivers the first of DOE’s 14-ton depleted uranium cylinders to USEC for re-enrichment as part of a five-party agreement that is extending enrichment operations at the 60-year-old plant for another year, delaying increased costs at the site for DOE. This cylinder hauler at Paducah's Babcock & Wilcox Conversion Services plant delivers the first of DOE's 14-ton depleted uranium cylinders to USEC for re-enrichment as part of a five-party agreement that is extending enrichment operations at the 60-year-old plant for another year, delaying

378

Measurements with an unreflected uranium (93.2%) metal sphere  

SciTech Connect

A near spherical unreflected and unmoderate uranium (93.7) metal configuration had been assembled to delayed criticality at Los Alamos National Laboratory in the 1950`s. Experiments with highly enriched uranium metal spherical shells had also been assembled. Both these experiments have been used to estimate the unreflected and unmoderated highly enriched uranium spherical critical mass. The experiments described in this paper, although originally justified for leakage spectra measurements and to investigate the use of a multiplying booster with a linear accelerator, also can provide estimates of the unreflected and unmoderated, highly enriched uranium spherical metal critical mass.

Mihalczo, J.T.; Lynn, J.J.; Taylor, J.R. [Oak Ridge National Lab., TN (United States); Hansen, G.E. [Los Alamos National Lab., NM (United States)

1993-06-01T23:59:59.000Z

379

Catalytic Self-Decontaminating Materials - Energy Innovation ...  

Self-decontaminating structures based on porphyrin-embedded, target imprinted, ... Biomass and Biofuels; Building Energy Efficiency; Electricity Transmission;

380

Chemical Warfare Agent Decontamination Foaming Composition And ...  

U.S. Energy Information Administration (EIA)

Field of the InventionThe present invention relates to foaming chemical warfare agent decontamination compositions. More particularly, ...

Note: This page contains sample records for the topic "uranium enrichment decontamination" 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

YNPS main coolant system decontamination  

SciTech Connect

The Yankee Nuclear Power Station (YNPS) located in Rowe, Massachusetts, is a four-loop pressurized water reactor that permanently ceased power operation on February 26, 1992. Decommissioning activities, including steam generator removal, reactor internals removal, and system dismantlement, have been in progress since the shutdown. One of the most significant challenges for YNPS in 1996 was the performance of the main coolant system chemical decontamination. This paper describes the objectives, challenges, and achievements involved in the planning and implementation of the chemical decontamination.

Metcalf, E.T. [Yankee Atomic Electric Co., Bolton, MA (United States)

1996-12-31T23:59:59.000Z

382

Uranium at Y-12: Accountability | Y-12 National Security Complex  

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

... ... Uranium at Y-12: Accountability Posted: July 22, 2013 - 3:37pm | Y-12 Report | Volume 10, Issue 1 | 2013 Accountability of enriched uranium is facilitated by the ability to put uranium into well-blended aqueous, organic, crystalline, powder, granular, metallic and compound forms that can be sampled and analyzed. Periodic inventories are necessary to find and account for all the enriched uranium that hides in equipment corners and crevices. This allows enriched uranium to be processed in large quantities and accounted for by the gram. Y-12 employees know where uranium resides in large, complex facilities and how to use computer tools to track and monitor its movement (see Uranium Track Team). Learn more about some of the complexities in reprocessing and safeguarding

383

Decontamination of steel by melt refining: A literature review  

SciTech Connect

It has been reported that a large amount of metal waste is produced annually by nuclear fuel processing and nuclear power plants. These metal wastes are contaminated with radioactive elements, such as uranium and plutonium. Current Department of Energy guidelines require retrievable storage of all metallic wastes containing transuranic elements above a certain level. Because of high cost, it is important to develop an effective decontamination and volume reduction method for low level contaminated metals. It has been shown by some investigators that a melt refining technique can be used for the processing of the contaminated metal wastes. In this process, contaminated metal is melted wit a suitable flux. The radioactive elements are oxidized and transferred to a slag phase. In order to develop a commercial process it is important to have information on the thermodynamics and kinetics of the removal. Therefore, a literature search was carried out to evaluate the available information on the decontamination uranium and transuranic-contaminated plain steel, copper and stainless steel by melt a refining technique. Emphasis was given to the thermodynamics and kinetics of the removal. Data published in the literature indicate that it is possible to reduce the concentration of radioactive elements to a very low level by the melt refining method. 20 refs.

Ozturk, B.; Fruehan, R.J. [Carnegie-Mellon Univ., Pittsburgh, PA (United States)

1994-12-31T23:59:59.000Z

384

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

385

Concrete Decontamination Technology Workshop Proceedings  

Science Conference Proceedings (OSTI)

EPRI has initiated a series of highly focused workshops, each dealing with a specific nuclear power plant decommissioning technology. The objective is to equip utility personnel with the information needed to assess the use of these technologies in their individual projects. This report presents the results of the first workshop, which covered concrete decontamination.

1998-12-31T23:59:59.000Z

386

Heterogeneous modeling of the uranium in situ recovery: Kinetic versus solubility Jrmy. Nosa,1, 2  

E-Print Network (OSTI)

Heterogeneous modeling of the uranium in situ recovery: Kinetic versus solubility control Jérémy Mines, Tour AREVA, 1 place Jean Millier, 92084 Paris La Défense Cedex, France The uranium in situ, into the deposit to selectively dissolve uranium. The solution enriched in uranium is pumped out and processed

Paris-Sud XI, Université de

387

Decontamination processes for waste glass canisters  

SciTech Connect

The process which will be used to decontaminate waste glass canisters at the Savannah River Plant consists of: decontamination (slurry blasting); rinse (high-pressure water); and spot decontamination (high-pressure water plus slurry). No additional waste will be produced by this process because glass frit used in decontamination will be mixed with the radioactive waste and fed into the glass melter. Decontamination of waste glass canisters with chemical and abrasive blasting techniques was investigated. The ability of a chemical technique with HNO/sub 3/-HF and H/sub 2/C/sub 2/O/sub 4/ to remove baked-on contamination was demonstrated. A correlation between oxide removal and decontamination was observed. Oxide removal and, thus, decontamination by abrasive blasting techniques with glass frit as the abrasive was proposed and demonstrated.

Rankin, W.N.

1981-06-01T23:59:59.000Z

388

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

389

URANIUM ALLOYS  

DOE Patents (OSTI)

A uranium alloy is reported containing from 0.1 to 5 per cent by weight of molybdenum and from 0.1 to 5 per cent by weight of silicon, the balance being uranium.

Colbeck, E.W.

1959-12-29T23:59:59.000Z

390

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

391

Laser Enrichment LLC Early Submittal of an Environmental Report,"  

E-Print Network (OSTI)

70.21, GE-Hitachi Global Laser Enrichment LLC (GLE) is submitting an application for the construction and operation of the GLE Commercial Facility in accordance with the requirements of 10 CFR Parts 30, 40, and 70. This proposed uranium enrichment facility will utilize a laser-based isotope separation technology to enrich uranium hexafluoride up to 8%, will have a nominal capacity of up to six million separative work units, and will be located in New Hanover County, North Carolina. On January 30, 2009, the GLE Commercial Facility Environmental Report was submitted

Tammy G. Orr; Michael F. Weber

2009-01-01T23:59:59.000Z

392

Overview of Depleted Uranium Hexafluoride Management Program  

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

DOE's DUF DOE's DUF 6 Cylinder Inventory a Location Number of Cylinders DUF 6 (MT) b Paducah, Kentucky 36,910 450,000 Portsmouth, Ohio 16,041 198,000 Oak Ridge (ETTP), Tennessee 4,683 56,000 Total 57,634 704,000 a The DOE inventory includes DUF 6 generated by the government, as well as DUF 6 transferred from U.S. Enrichment Corporation pursuant to two memoranda of agreement. b A metric ton (MT) is equal to 1,000 kilograms, or 2,200 pounds. Overview of Depleted Uranium Hexafluoride Management Program Over the last four decades, large quantities of uranium were processed by gaseous diffusion to produce enriched uranium for U.S. national defense and civilian purposes. The gaseous diffusion process uses uranium in the form of uranium hexafluoride (UF 6 ), primarily because UF 6 can conveniently be used in

393

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

394

Pyrolitic Uranium Compound (PYRUC)  

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

Pyrolitic Uranium Compound Pyrolitic Uranium Compound (PYRUC) PYRolitic Uranium Compound (PYRUC) is a shielding material consisting of depleted uranium UO2 or UC in either pellet...

395

Contaminant-Organic Complexes: Their Structure and Energetics in Surface Decontamination Processes  

SciTech Connect

The current debate over possible decontamination processes for DOE facilities is centered on disparate decontamination problems, but the key contaminants (Thorium [Th],uranium [U], and plutonium [Pu]) are universally important. Innovative agents used alone or in conjunction with traditional processes can increase the potential to reclaim for future use some these valuable resources or at the least decontaminate the metal surfaces to allow disposal as nonradioactive, nonhazardous material. This debate underscores several important issues: (1) regardless of the decontamination scenario, metal (Fe, U, Pu, Np) oxide film removal from the surface is central to decontamination; and (2) simultaneous oxide dissolution and sequestration of actinide contaminants against re-adsorption to a clean metal surface will influence the efficacy of a process or agent and its cost. Current research is investigating the use of microbial siderophores (chelates) to solubilize actinides (i.e., Th, U, Pu) from the surface of Fe oxide surfaces. Continuing research integrates (1) studies of macroscopic dissolution/desorption of common actinide (IV) [Th, U, Pu, Np] solids and species sorbed to and incorporated into Fe oxides, (2) molecular spectroscopy (FTIR, Raman, XAS), to probe the structure and bonding of contaminants, siderophores and their functional moieties, and how these change with the chemical environment, (3) and molecular mechanics and electronic structure calculations to design model siderophore compounds to test and extend the MM3 model.

Ainsworth, Calvin C.; Hay, Benjamin P.; Traina, Samuel J.; Myneni, Satish C. B.

2003-06-01T23:59:59.000Z

396

Energy Department Selects Global Laser Enrichment for Future Operations at  

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

Energy Department Selects Global Laser Enrichment for Future Energy Department Selects Global Laser Enrichment for Future Operations at Paducah Site Energy Department Selects Global Laser Enrichment for Future Operations at Paducah Site November 27, 2013 - 12:00pm Addthis Workers inspect cylinders containing depleted uranium hexafluoride. Workers inspect cylinders containing depleted uranium hexafluoride. Media Contact (202) 586-4940 Washington, D.C. - The U.S. Department of Energy announced today that it will open negotiations with Global Laser Enrichment (GLE) for the sale of the depleted uranium hexafluoride inventory. The Department determined that GLE offered the greatest benefit to the government among those who responded to a Request for Offers (RFO) released earlier this year. Through the RFO review process, the Department also decided to enter into

397

New Waste Calcining Facility Non-radioactive Process Decontamination  

Science Conference Proceedings (OSTI)

This report documents the results of a test of the New Calcining Facility (NWCF) process decontamination system. The decontamination system test occurred in December 1981, during non-radioactive testing of the NWCF. The purpose of the decontamination system test was to identify equipment whose design prevented effective calcine removal and decontamination. Effective equipment decontamination was essential to reduce radiation fields for in-cell work after radioactive processing began. The decontamination system test began with a pre-decontamination inspection of the equipment. The pre-decontamination inspection documented the initial condition and cleanliness of the equipment. It provided a basis for judging the effectiveness of the decontamination. The decontamination consisted of a series of equipment flushes using nitric acid and water. A post-decontamination equipment inspection determined the effectiveness of the decontamination. The pre-decontamination and post-decontamination equipment inspections were documented with hotographs. The decontamination system was effective in removing calcine from most of the NWCF equipment as evidenced by little visible calcine residue in the equipment after decontamination. The decontamination test identified four areas where the decontamination system required improvement. These included the Calciner off-gas line, Cyclone off-gas line, fluidizing air line, and the Calciner baffle plates. Physical modifications to enhance decontamination were made to those areas, resulting in an effective NWCF decontamination system.

Swenson, Michael Clair

2001-09-01T23:59:59.000Z

398

New Waste Calcining Facility Non-Radioactive Process Decontamination  

SciTech Connect

This report documents the results of a test of the New Calcining Facility (NWCF) process decontamination system. The decontamination system test occurred in December 1981, during non-radioactive testing of the NWCF. The purpose of the decontamination system test was to identify equipment whose design prevented effective calcine removal and decontamination. Effective equipment decontamination was essential to reduce radiation fields for in-cell work after radioactive processing began. The decontamination system test began with a pre-decontamination inspection of the equipment. The pre- decontamination inspection documented the initial condition and cleanliness of the equipment. It provided a basis for judging the effectiveness of the decontamination. The decontamination consisted of a series of equipment flushes using nitric acid and water. A post-decontamination equipment inspection determined the effectiveness of the decontamination. The pre-decontamination and post-decontamination equipment inspections were documented with photographs. The decontamination system was effective in removing calcine from most of the NWCF equipment as evidenced by little visible calcine residue in the equipment after decontamination. The decontamination test identified four areas where the decontamination system required improvement. These included the Calciner off-gas line, Cyclone off-gas line, fluidizing air line, and the Calciner baffle plates. Physical modifications to enhance decontamination were made to those areas, resulting in an effective NWCF decontamination system.

Swenson, Michael C.

2001-09-30T23:59:59.000Z

399

D&D of the French High Enrichment Gaseous Diffusion Plant  

SciTech Connect

This paper describes the D&D program that is being implemented at France's High Enrichment Gaseous Diffusion Plant, which was designed to supply France's Military with Highly Enriched Uranium. This plant was definitively shut down in June 1996, following French President Jacques Chirac's decision to end production of Highly Enriched Uranium and dismantle the corresponding facilities.

BEHAR, Christophe; GUIBERTEAU, Philippe; DUPERRET, Bernard; TAUZIN, Claude

2003-02-27T23:59:59.000Z

400

U.S. Uranium Down-blending Activities: Fact Sheet | National...  

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

Down-blending Activities: Fact Sheet Mar 23, 2012 The permanent disposition of Highly Enriched Uranium (HEU) permanently reduces nuclear security vulnerabilities. In 1996, the...

Note: This page contains sample records for the topic "uranium enrichment decontamination" 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.


401

Reductive dissolution approaches to removal of uranium from contaminated soils  

SciTech Connect

Traditional approaches to uranium recovery from ores have employed oxidation of U(IV) minerals to form the uranyl cation which is subsequently complexed by carbonate or maintained in solution by strong acids. Reductive approaches for uranium decontamination have been limited to removing soluble uranium from solutions by formation of U{sup 4+} which readily hydrolyses and precipitates. As part of the Uranium in Soils Integrated Demonstration, we have developed a reductive approach to solubilization of uranium from contaminated soils which employs reduction to destabilize U(VI) solid and sorbed species, and strong chelators for U(IV) to prevent hydrolysis and solubilize the reduced from. This strategy has particular application to sites where the uranium is present primarily as intractable U(VI) phases and where high fractions of the contamination must be removed to meet regulatory requirements.

Brainard, J.R.; Iams, H.D.; Strietelmeier, B.A.; Del-Rio Garcia, M.

1994-06-01T23:59:59.000Z

402

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

403

Proceedings: 1998 EPRI Chemical Decontamination Conference  

Science Conference Proceedings (OSTI)

In today's competitive environment, chemical decontamination technology has evolved to meet nuclear industry challenges to control costs and reduce outage times. EPRI's 1998 Chemical Decontamination Conference included 29 presentations which highlighted recent technology developments and field experience. Featured in the conference were the results of the first applications of the EPRI decontamination for decommissioning (DFD) process in the decommissioning of Big Rock Point and Maine Yankee plants.

1998-07-29T23:59:59.000Z

404

Predicting 232U Content in Uranium  

SciTech Connect

The minor isotope 232U may ultimately be used for detection or confirmation of uranium in a variety of applications. The primary advantage of 232 U as an indicator of the presence of enriched uranium is the plentiful and penetrating nature of the radiation emitted by its daughter radionuclide 208Tl. A possible drawback to measuring uranium via 232U is the relatively high uncertainty in 232U abundance both within and between material populations. An important step in assessing this problem is to ascertain what determines the 232U concentration within any particular sample of uranium. To this end, we here analyze the production and eventual enrichment of 232 U during fuel-cycle operations. The goal of this analysis is to allow approximate prediction of 232 U quantities, or at least some interpretation of the results of 232U measurements. We have found that 232U is produced via a number of pathways during reactor irradiation of uranium and is subsequently concentrated during the later enrichment of the uranium' s 235U Content. While exact calculations are nearly impossible for both the reactor-production and cascade-enrichment parts of the prediction problem, estimates and physical bounds can be provided as listed below and detailed within the body of the report. Even if precise calculations for the irradiation and enrichment were possible, the ultimate 212U concentration would still depend upon the detailed fuel-cycle history. Assuming that a thennal-diffusion cascade is used to produce highly enriched uranium (HEU), dilution of reactor-processed fuel at the cascade input and the long-term holdup of 232U within the cascade both affect the 232U concentration in the product. Similar issues could be expected to apply for the other isotope-separation technologies that are used in other countries. Results of this analysis are listed below: 0 The 232U concentration depends strongly on the uranium enrichment, with depleted uranium (DU) containing between 1600 and 8000 times less 232U than HEU does. * The 236U/232U concentration ratio in HEU is likely to be between 10{sup 6} and 2 x 10{sup 7}. 0 Plutonium-production reactors yield uranium with between I and 10 ppt of 232u. 0 Much higher 132U concentrations can be obtained in some situations. * Significant variation in the 232U concentration is inevitable. * Cascade enrichment increases the 232U concentration by a factor of at least 200, and possibly as much as 1000. 0 The actual 232U concentration depends upon the dilution at the cascade input.

AJ Peurrung

1999-01-07T23:59:59.000Z

405

Predicting 232U Content in Uranium  

SciTech Connect

The minor isotope 232U may ultimately be used for detection or confirmation of uranium in a variety of applications. The primary advantage of 232 U as an indicator of the presence of enriched uranium is the plentiful and penetrating nature of the radiation emitted by its daughter radionuclide 208Tl. A possible drawback to measuring uranium via 232U is the relatively high uncertainty in 232U abundance both within and between material populations. An important step in assessing this problem is to ascertain what determines the 232U concentration within any particular sample of uranium. To this end, we here analyze the production and eventual enrichment of 232 U during fuel-cycle operations. The goal of this analysis is to allow approximate prediction of 232 U quantities, or at least some interpretation of the results of 232U measurements. We have found that 232U is produced via a number of pathways during reactor irradiation of uranium and is subsequently concentrated during the later enrichment of the uranium' s 235U Content. While exact calculations are nearly impossible for both the reactor-production and cascade-enrichment parts of the prediction problem, estimates and physical bounds can be provided as listed below and detailed within the body of the report. Even if precise calculations for the irradiation and enrichment were possible, the ultimate 212U concentration would still depend upon the detailed fuel-cycle history. Assuming that a thennal-diffusion cascade is used to produce highly enriched uranium (HEU), dilution of reactor-processed fuel at the cascade input and the long-term holdup of 232U within the cascade both affect the 232U concentration in the product. Similar issues could be expected to apply for the other isotope-separation technologies that are used in other countries. Results of this analysis are listed below: 0 The 232U concentration depends strongly on the uranium enrichment, with depleted uranium (DU) containing between 1600 and 8000 times less 232U than HEU does. * The 236U/232U concentration ratio in HEU is likely to be between 10{sup 6} and 2 x 10{sup 7}. 0 Plutonium-production reactors yield uranium with between I and 10 ppt of 232u. 0 Much higher 132U concentrations can be obtained in some situations. * Significant variation in the 232U concentration is inevitable. * Cascade enrichment increases the 232U concentration by a factor of at least 200, and possibly as much as 1000. 0 The actual 232U concentration depends upon the dilution at the cascade input.

AJ Peurrung

1999-01-07T23:59:59.000Z

406

Novel Functionalized Nanomaterials for Organic Decontamination  

Science Conference Proceedings (OSTI)

... oxide (TiO2)-graphene-rhamnolipid for decontamination of organics (methyl orange, phenol and diesel) from water. The results show the advantage of organic...

407

Micro Aerosol-based Decontamination System - Available ...  

Search PNNL. PNNL Home; About; Research; Publications; Jobs; News; Contacts; Micro Aerosol-based Decontamination System. Battelle Number(s): 15847. ...

408

Vibratory finishing as a decontamination process  

SciTech Connect

The major objective of this research is to develop vibratory finishing into a large-scale decontamination technique that can economicaly remove transuranic and other surface contamination from large volumes of waste produced by the operation and decommissioning of retired nuclear facilities. The successful development and widespread application of this decontamination technique would substantially reduce the volume of waste requiring expensive geologic disposal. Other benefits include exposure reduction for decontamination personnel and reduced risk of environmental contamination. Laboratory-scale studies showed that vibratory finishing can rapidly reduce the contamination level of transuranic-contaminated stainless steel and Plexiglas to well below the 10-nCi/g limit. The capability of vibratory finishing as a decontamination process was demonstrated on a large scale. The first decontamination demonstration was conducted at the Hanford N-Reactor, where a vibratory finisher was installed to reduce personnel exposure during the summer outage. Items decontaminated included fuel spacers, process-tube end caps, process-tube inserts, pump parts, ball-channel inspection tools and miscellaneous hand tools. A second demonstration is currently being conducted in the decontamination facility at the Hanford 231-Z Building. During this demonstration, transuranic-contaminated material from decommissioned plutonium facilities is being decontaminated to <10 nCi/g to minimize the volume of material that will require geologic disposal. Items that are being decontaminated include entire glove boxes, process-hood structural material and panels, process tanks, process-tank shields, pumps, valves and hand tools used during the decommissioning work.

McCoy, M.W.; Arrowsmith, H.W.; Allen, R.P.

1980-10-01T23:59:59.000Z

409

Radioactive Material or Multiple Hazardous Materials Decontamination  

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

The purpose of this procedure is to provide guidance for performing decontamination ofindividuals who have entered a hot zone during transportation incidents involving radioactive.

410

Decontamination of Biological Threats in Water Supplies  

Science Conference Proceedings (OSTI)

Decontamination of Biological Threats in Water Supplies. ... The availability of safe pure drinking water in the United States is taken for granted. ...

2012-10-01T23:59:59.000Z

411

Chemical Warfare Agent Decontamination Solution - Patent 5859064  

U.S. Energy Information Administration (EIA)

... and softens and removes paint.A need exists for a chemical warfare agent decontamination solution which is noncorrosive, nontoxic, nonflammable, ...

412

Chemical Agent Decontamination Composition Comprising A ...  

U.S. Energy Information Administration (EIA)

Chemical warfare agents are stockpiled ... but also in today's climate of terrorist threats of WMD chemical attacks.Methods for decontamination of che ...

413

Proceedings of the concrete decontamination workshop  

Science Conference Proceedings (OSTI)

Fourteen papers were presented. These papers describe concrete surface removal methods and equipment, as well as experiences in decontaminating and removing both power and experimental nuclear reactors.

Halter, J.M.; Sullivan, R.G.; Currier, A.J.

1980-05-28T23:59:59.000Z

414

Domestic utility attitudes toward foreign uranium supply  

SciTech Connect

The current embargo on the enrichment of foreign-origin uranium for use in domestic utilization facilities is scheduled to be removed in 1984. The pending removal of this embargo, complicated by a depressed worldwide market for uranium, has prompted consideration of a new or extended embargo within the US Government. As part of its on-going data collection activities, Nuclear Resources International (NRI) has surveyed 50 domestic utility/utility holding companies (representing 60 lead operator-utilities) on their foreign uranium purchase strategies and intentions. The most recent survey was conducted in early May 1981. A number of qualitative observations were made during the course of the survey. The major observations are: domestic utility views toward foreign uranium purchase are dynamic; all but three utilities had some considered foreign purchase strategy; some utilities have problems with buying foreign uranium from particular countries; an inducement is often required by some utilities to buy foreign uranium; opinions varied among utilities concerning the viability of the domestic uranium industry; and many utilities could have foreign uranium fed through their domestic uranium contracts (indirect purchases). The above observations are expanded in the final section of the report. However, it should be noted that two of the observations are particularly important and should be seriously considered in formulation of foreign uranium import restrictions. These important observations are the dynamic nature of the subject matter and the potentially large and imbalanced effect the indirect purchases could have on utility foreign uranium procurement.

1981-06-01T23:59:59.000Z

415

Uranium Processing Facility | Y-12 National Security Complex  

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

About / Transforming Y-12 / Uranium Processing Facility About / Transforming Y-12 / Uranium Processing Facility Uranium Processing Facility UPF will be a state-of-the-art, consolidated facility for enriched uranium operations including assembly, disassembly, dismantlement, quality evaluation, and product certification. An integral part of Y-12's transformation efforts and a key component of the National Nuclear Security Administration's Uranium Center of Excellence, the Uranium Processing Facility is one of two facilities at Y-12 whose joint mission will be to accomplish the storage and processing of all enriched uranium in one much smaller, centralized area. Safety, security and flexibility are key design attributes of the facility, which is in the preliminary design phase of work. UPF will be built to modern standards and engage new technologies through a responsive and agile

416

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,

417

Selective leaching of uranium from uranium-contaminated soils  

SciTech Connect

Three soils and a sediment contaminated with uranium were used to determine the effectiveness of sodium carbonate and citric acid leaching to decontaminate or remove uranium to acceptable regulatory levels. The objective was to selectively extract uranium using a soil washing/extraction process without seriously degrading the soil`s physicochemical characteristics or generating a secondary waste form that would be difficult to manage and/or dispose of. Two of the soils were surface soils from the DOE facility formerly called the Feed Materials Production Center (FMPC) at Fernald, Ohio. One of the soils is from near the Plant 1 storage pad and the other soil was taken from near a waste incinerator used to burn low-level contaminated trash. The third soil was a surface soil from an area formally used as a landfarm for the treatment of spent oils at the Oak Ridge Y-12 Plant. The sediment sample was material sampled from a storm sewer sediment trap at the Oak Ridge Y-12 Plant. Uranium concentrations in the Fernald soils ranged from 450 to 550 {mu}g U/g of soil while the samples from the Y-12 Plant ranged from 150 to 200 {mu}g U/g of soil.

Francis, C.W.; Mattus, A.J.; Farr, L.L.; Lee, S.Y. [Oak Ridge National Lab., TN (United States); Elless, M.P. [Oak Ridge National Lab., TN (United States)]|[Oak Ridge Associated Universities, Inc., TN (United States)

1993-06-01T23:59:59.000Z

418

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

419

Anthrax Sampling and Decontamination: Technology Trade-Offs  

E-Print Network (OSTI)

be sampled; (c) the cost per square foot to decontaminate acritical items. The cost per square foot of decontaminatingcritical items. The cost per square foot of decontaminating

Price, Phillip N.

2009-01-01T23:59:59.000Z

420

EA-1266: Proposed Decontamination and Disassembly of the Argonne...  

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

66: Proposed Decontamination and Disassembly of the Argonne Thermal Source Reactor (ATSR) At Argonne National Laboratory, Argonne, Illinois EA-1266: Proposed Decontamination and...

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