31 CFR 540.306 - Highly Enriched Uranium (HEU).

Code of Federal Regulations, 2013 CFR

2013-07-01

... 31 Money and Finance:Treasury 3 2013-07-01 2013-07-01 false Highly Enriched Uranium (HEU). 540.306... OF FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.306 Highly Enriched Uranium (HEU). The term highly enriched...

31 CFR 540.308 - Low Enriched Uranium (LEU).

Code of Federal Regulations, 2014 CFR

2014-07-01

... 31 Money and Finance:Treasury 3 2014-07-01 2014-07-01 false Low Enriched Uranium (LEU). 540.308... OF FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.308 Low Enriched Uranium (LEU). The term low enriched...

31 CFR 540.306 - Highly Enriched Uranium (HEU).

Code of Federal Regulations, 2014 CFR

2014-07-01

... 31 Money and Finance:Treasury 3 2014-07-01 2014-07-01 false Highly Enriched Uranium (HEU). 540.306... OF FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.306 Highly Enriched Uranium (HEU). The term highly enriched...

31 CFR 540.308 - Low Enriched Uranium (LEU).

Code of Federal Regulations, 2011 CFR

2011-07-01

... 31 Money and Finance:Treasury 3 2011-07-01 2011-07-01 false Low Enriched Uranium (LEU). 540.308... OF FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.308 Low Enriched Uranium (LEU). The term low enriched...

31 CFR 540.306 - Highly Enriched Uranium (HEU).

Code of Federal Regulations, 2012 CFR

2012-07-01

... 31 Money and Finance:Treasury 3 2012-07-01 2012-07-01 false Highly Enriched Uranium (HEU). 540.306... OF FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.306 Highly Enriched Uranium (HEU). The term highly enriched...

31 CFR 540.308 - Low Enriched Uranium (LEU).

Code of Federal Regulations, 2012 CFR

2012-07-01

... 31 Money and Finance:Treasury 3 2012-07-01 2012-07-01 false Low Enriched Uranium (LEU). 540.308... OF FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.308 Low Enriched Uranium (LEU). The term low enriched...

31 CFR 540.308 - Low Enriched Uranium (LEU).

Code of Federal Regulations, 2013 CFR

2013-07-01

... 31 Money and Finance:Treasury 3 2013-07-01 2013-07-01 false Low Enriched Uranium (LEU). 540.308... OF FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.308 Low Enriched Uranium (LEU). The term low enriched...

31 CFR 540.308 - Low Enriched Uranium (LEU).

Code of Federal Regulations, 2010 CFR

2010-07-01

... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false Low Enriched Uranium (LEU). 540.308... OF FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.308 Low Enriched Uranium (LEU). The term low enriched...

Apparatus to recover tritium from tritiated molecules

DOEpatents

Swansiger, William A.

1988-01-01

An apparatus for recovering tritium from tritiated compounds is provided, including a preheater for heating tritiated water and other co-injected tritiated compounds to temperatures of about 600.degree. C. and a reactor charged with a mixture of uranium and uranium dioxide for receiving the preheated mixture. The reactor vessel is preferably stainless steel of sufficient mass so as to function as a heat sink preventing the reactor side walls from approaching high temperatures. A disposable copper liner extends between the reaction chamber and stainless steel outer vessel to prevent alloying of the uranium with the outer vessel. The uranium dioxide functions as an insulating material and heat sink preventing the reactor side walls from attaining reaction temperatures to thereby minimize tritium permeation rates. The uranium dioxide also functions as a diluent to allow for volumetric expansion of the uranium as it is converted to uranium dioxide.

Laser and gas centrifuge enrichment

NASA Astrophysics Data System (ADS)

Heinonen, Olli

2014-05-01

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2012-08-24

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

31 CFR 540.316 - Uranium enrichment.

Code of Federal Regulations, 2010 CFR

2010-07-01

... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false Uranium enrichment. 540.316 Section 540.316 Money and Finance: Treasury Regulations Relating to Money and Finance (Continued) OFFICE OF... REGULATIONS General Definitions § 540.316 Uranium enrichment. The term uranium enrichment means the process of...

31 CFR 540.306 - Highly Enriched Uranium (HEU).

Code of Federal Regulations, 2010 CFR

2010-07-01

... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false Highly Enriched Uranium (HEU). 540...) OFFICE OF FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.306 Highly Enriched Uranium (HEU). The term highly...

10 CFR 70.23a - Hearing required for uranium enrichment facility.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Hearing required for uranium enrichment facility. 70.23a... MATERIAL License Applications § 70.23a Hearing required for uranium enrichment facility. The Commission... license for construction and operation of a uranium enrichment facility. The Commission will publish...

10 CFR 70.23a - Hearing required for uranium enrichment facility.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 2 2012-01-01 2012-01-01 false Hearing required for uranium enrichment facility. 70.23a... MATERIAL License Applications § 70.23a Hearing required for uranium enrichment facility. The Commission... license for construction and operation of a uranium enrichment facility. The Commission will publish...

10 CFR 70.23a - Hearing required for uranium enrichment facility.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 2 2013-01-01 2013-01-01 false Hearing required for uranium enrichment facility. 70.23a... MATERIAL License Applications § 70.23a Hearing required for uranium enrichment facility. The Commission... license for construction and operation of a uranium enrichment facility. The Commission will publish...

10 CFR 70.23a - Hearing required for uranium enrichment facility.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 2 2014-01-01 2014-01-01 false Hearing required for uranium enrichment facility. 70.23a... MATERIAL License Applications § 70.23a Hearing required for uranium enrichment facility. The Commission... license for construction and operation of a uranium enrichment facility. The Commission will publish...

10 CFR 70.23a - Hearing required for uranium enrichment facility.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Hearing required for uranium enrichment facility. 70.23a... MATERIAL License Applications § 70.23a Hearing required for uranium enrichment facility. The Commission... license for construction and operation of a uranium enrichment facility. The Commission will publish...

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

Federal Register 2010, 2011, 2012, 2013, 2014

2011-11-02

... Uranium Enrichment Fuel Cycle Facility's Inspection Reports Regarding Louisiana Energy Services, National..., Uranium Enrichment Branch, Division of Fuel Cycle Safety and Safeguards, Office of Nuclear Material Safety... Commission. Brian W. Smith, Chief, Uranium Enrichment Branch, Division of Fuel Cycle Safety and Safeguards...

78 FR 77650 - Low Enriched Uranium From France: Continuation of Antidumping Duty Order

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-24

... DEPARTMENT OF COMMERCE International Trade Administration [A-427-818] Low Enriched Uranium From... Commission (the ``ITC'') that revocation of the antidumping duty order on low enriched uranium (``LEU'') from... Initiation of Five-Year (``Sunset'') Review, 77 FR 71684 (December 3, 2013). \\2\\ See Low Enriched Uranium...

78 FR 21416 - Low Enriched Uranium From France; Scheduling of a Full Five-year Review Concerning the...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-04-10

... INTERNATIONAL TRADE COMMISSION [Investigation No. 731-TA-909 (Second Review)] Low Enriched Uranium... Enriched Uranium from France AGENCY: United States International Trade Commission. ACTION: Notice. SUMMARY... antidumping duty order on low enriched uranium from France would be likely to lead to continuation or...

Assuaging Nuclear Energy Risks: The Angarsk International Uranium Enrichment Center

NASA Astrophysics Data System (ADS)

Myers, Astasia

2011-06-01

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.

10 CFR 40.33 - Issuance of a license for a uranium enrichment facility.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 1 2011-01-01 2011-01-01 false Issuance of a license for a uranium enrichment facility... License Applications § 40.33 Issuance of a license for a uranium enrichment facility. (a) The Commission... the licensing of the construction and operation of a uranium enrichment facility. The Commission will...

10 CFR 40.33 - Issuance of a license for a uranium enrichment facility.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 1 2012-01-01 2012-01-01 false Issuance of a license for a uranium enrichment facility... License Applications § 40.33 Issuance of a license for a uranium enrichment facility. (a) The Commission... the licensing of the construction and operation of a uranium enrichment facility. The Commission will...

10 CFR 40.33 - Issuance of a license for a uranium enrichment facility.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 1 2013-01-01 2013-01-01 false Issuance of a license for a uranium enrichment facility... License Applications § 40.33 Issuance of a license for a uranium enrichment facility. (a) The Commission... the licensing of the construction and operation of a uranium enrichment facility. The Commission will...

10 CFR 40.33 - Issuance of a license for a uranium enrichment facility.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 1 2014-01-01 2014-01-01 false Issuance of a license for a uranium enrichment facility... License Applications § 40.33 Issuance of a license for a uranium enrichment facility. (a) The Commission... the licensing of the construction and operation of a uranium enrichment facility. The Commission will...

Recovery of tritium from tritiated molecules

DOEpatents

Swansiger, William A.

1987-01-01

A method of recovering tritium from tritiated compounds comprises the steps of heating tritiated water and other co-injected tritiated compounds in a preheater to temperatures of about 600.degree. C. The mixture is injected into a reactor charged with a mixture of uranium and uranium dioxide. The injected mixture undergoes highly exothermic reactions with the uranium causing reaction temperatures to occur in excess of the melting point of uranium, and complete decomposition of the tritiated compounds to remove tritium therefrom. The uranium dioxide functions as an insulating material and heat sink preventing the reactor side walls from attaining reaction temperatures to thereby minimize tritium permeation rates. The uranium dioxide also functions as a diluent to allow for volumetric expansion of the uranium as it is converted to uranium dioxide. The reactor vessel is preferably stainless steel of sufficient mass so as to function as a heat sink preventing the reactor side walls from approaching high temperatures. A disposable copper liner extends between the reaction chamber and stainless steel outer vessel to prevent alloying of the uranium with the outer vessel. Apparatus used to carry out the method of the invention is also disclosed.

DECONTAMINATION OF URANIUM

DOEpatents

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

1958-02-01

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.

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

DOEpatents

Horton, James A.; Hayden, Jr., Howard W.

1995-01-01

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.

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

DOEpatents

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

1995-05-30

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.

MANUFACTURE OF UF$sub 4$

DOEpatents

Calcott, W.S.

1959-10-13

The manufacture of uranium tetrafluoride from urarium dioxide is described. Uranium dioxide is heated to about 500 deg C in a reactor. Anhydrous hydrogen fluoride is passed through the reactor in contact with uranium dioxide for several hours, the flow of hydrogen fluoride is discontinued, and hydrogen passed through the reactor for less than an hour. The flow of hydrogen fluoride is resumed for several hours, and then nitrogen is passed for a few minutes to expel unreacted hydrogen fluoride as water vapor. The reactor is cooled to room temperature and the uranium tetrafluoride removed.

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

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-30

... NUCLEAR REGULATORY COMMISSION [Docket No. 70-3103; NRC-2010-0264] Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding Louisiana Energy Services LLC, National Enrichment Facility, Eunice..., Chief, Uranium Enrichment Branch, Division of Fuel Cycle Safety and Safeguards, Office of Nuclear...

High Temperature Reactions of Uranium Dioxide with Various Metal Oxides

DTIC Science & Technology

1956-02-20

manganese, nickel , lead, and tin. Subtracting the total of these impurities from the oxygen remainder would give a more nearly 1:2 uranium -oxygen ratio. The...Astin, Dire~ctor High -Temperature Reactions of Uranium Dioxide With Various Metal Oxides Acceson For NTIS CRAWI DTfC TAB Unannounced D JustifiCation...1 2. The uranium -oxygen system ------------------------------------- 1 3. Binary systems containing

Net energy payback and CO2 emissions from three midwestern wind farms: An update

USGS Publications Warehouse

White, S.W.

2006-01-01

This paper updates a life-cycle net energy analysis and carbon dioxide emissions analysis of three Midwestern utility-scale wind systems. Both the Energy Payback Ratio (EPR) and CO2 analysis results provide useful data for policy discussions regarding an efficient and low-carbon energy mix. The EPR is the amount of electrical energy produced for the lifetime of the power plant divided by the total amount of energy required to procure and transport the materials, build, operate, and decommission the power plants. The CO2 analysis for each power plant was calculated from the life-cycle energy input data. A previous study also analyzed coal and nuclear fission power plants. At the time of that study, two of the three wind systems had less than a full year of generation data to project the life-cycle energy production. This study updates the analysis of three wind systems with an additional four to eight years of operating data. The EPR for the utility-scale wind systems ranges from a low of 11 for a two-turbine system in Wisconsin to 28 for a 143-turbine system in southwestern Minnesota. The EPR is 11 for coal, 25 for fission with gas centrifuge enriched uranium and 7 for gaseous diffusion enriched uranium. The normalized CO2 emissions, in tonnes of CO2 per GW eh, ranges from 14 to 33 for the wind systems, 974 for coal, and 10 and 34 for nuclear fission using gas centrifuge and gaseous diffusion enriched uranium, respectively. ?? Springer Science+Business Media, LLC 2007.

Feasibility study of a small, thorium-based fission power system for space and terrestrial applications

NASA Astrophysics Data System (ADS)

Worrall, Michael Jason

One of the current challenges facing space exploration is the creation of a power source capable of providing useful energy for the entire duration of a mission. Historically, radioisotope batteries have been used to provide load power, but this conventional system may not be capable of sustaining continuous power for longer duration missions. To remedy this, many forays into nuclear powered spacecraft have been investigated, but no robust system for long-term power generation has been found. In this study, a novel spin on the traditional fission power system that represents a potential optimum solution is presented. By utilizing mature High Temperature Gas Reactor (HTGR) technology in conjunction with the capabilities of the thorium fuel cycle, we have created a light-weight, long-term power source capable of a continuous electric power output of up to 70kW for over 15 years. This system relies upon a combination of fissile, highly-enriched uranium dioxide and fertile thorium carbide Tri-Structural Isotropic (TRISO) fuel particles embedded in a hexagonal beryllium oxide matrix. As the primary fissile material is consumed, the fertile material breeds new fissile material leading to more steady fuel loading over the lifetime of the core. Reactor control is achieved through an innovative approach to the conventional boron carbide neutron absorber by utilizing sections of borated aluminum placed in rotating control drums within the reflector. Borated aluminum allows for much smaller boron concentrations, thus eliminating the potential for 10B(n,alpha)6Li heating issues that are common in boron carbide systems. A wide range of other reactivity control systems are also investigated, such as a radially-split rotating reflector. Lastly, an extension of the design to a terrestrial based system is investigated. In this system, uranium enrichment is dropped to 20 percent in order to meet current regulations, a solid uranium-zirconium hydride fissile driver replaces the uranium dioxide TRISO particles, and the moderating material is changed from beryllium oxide to graphite. These changes result in an increased core size, but the same long-term power generation potential is achieved. Additionally, small amounts of erbium are added to the hydride matrix to further extend core lifetime.

10 CFR 140.13b - Amount of liability insurance required for uranium enrichment facilities.

Code of Federal Regulations, 2010 CFR

2010-01-01

... enrichment facilities. 140.13b Section 140.13b Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) FINANCIAL... required for uranium enrichment facilities. Each holder of a license issued under Parts 40 or 70 of this chapter for a uranium enrichment facility that involves the use of source material or special nuclear...

78 FR 37925 - Continuation of the National Emergency With Respect to the Disposition of Russian Highly Enriched...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-06-24

... National Emergency With Respect to the Disposition of Russian Highly Enriched Uranium On June 25, 2012, by... America and the Government of the Russian Federation Concerning the Disposition of Highly Enriched Uranium... Russian highly enriched uranium declared in Executive Order 13617. [[Page 37926

10 CFR 140.13b - Amount of liability insurance required for uranium enrichment facilities.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Amount of liability insurance required for uranium... required for uranium enrichment facilities. Each holder of a license issued under Parts 40 or 70 of this chapter for a uranium enrichment facility that involves the use of source material or special nuclear...

10 CFR 140.13b - Amount of liability insurance required for uranium enrichment facilities.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 2 2012-01-01 2012-01-01 false Amount of liability insurance required for uranium... required for uranium enrichment facilities. Each holder of a license issued under Parts 40 or 70 of this chapter for a uranium enrichment facility that involves the use of source material or special nuclear...

10 CFR 140.13b - Amount of liability insurance required for uranium enrichment facilities.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 2 2013-01-01 2013-01-01 false Amount of liability insurance required for uranium... required for uranium enrichment facilities. Each holder of a license issued under Parts 40 or 70 of this chapter for a uranium enrichment facility that involves the use of source material or special nuclear...

10 CFR 140.13b - Amount of liability insurance required for uranium enrichment facilities.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 2 2014-01-01 2014-01-01 false Amount of liability insurance required for uranium... required for uranium enrichment facilities. Each holder of a license issued under Parts 40 or 70 of this chapter for a uranium enrichment facility that involves the use of source material or special nuclear...

10 CFR 766.3 - Definitions.

Code of Federal Regulations, 2014 CFR

2014-01-01

... ENERGY URANIUM ENRICHMENT DECONTAMINATION AND DECOMMISSIONING FUND; PROCEDURES FOR SPECIAL ASSESSMENT OF... account in the U.S. Treasury referred to as the Uranium Enrichment Decontamination and Decommissioning... separative work unit, the common measure by which uranium enrichment services are sold. TESS means the Toll...

10 CFR 766.3 - Definitions.

Code of Federal Regulations, 2012 CFR

2012-01-01

... ENERGY URANIUM ENRICHMENT DECONTAMINATION AND DECOMMISSIONING FUND; PROCEDURES FOR SPECIAL ASSESSMENT OF... account in the U.S. Treasury referred to as the Uranium Enrichment Decontamination and Decommissioning... separative work unit, the common measure by which uranium enrichment services are sold. TESS means the Toll...

10 CFR 766.3 - Definitions.

Code of Federal Regulations, 2013 CFR

2013-01-01

... ENERGY URANIUM ENRICHMENT DECONTAMINATION AND DECOMMISSIONING FUND; PROCEDURES FOR SPECIAL ASSESSMENT OF... account in the U.S. Treasury referred to as the Uranium Enrichment Decontamination and Decommissioning... separative work unit, the common measure by which uranium enrichment services are sold. TESS means the Toll...

10 CFR 766.3 - Definitions.

Code of Federal Regulations, 2011 CFR

2011-01-01

... ENERGY URANIUM ENRICHMENT DECONTAMINATION AND DECOMMISSIONING FUND; PROCEDURES FOR SPECIAL ASSESSMENT OF... account in the U.S. Treasury referred to as the Uranium Enrichment Decontamination and Decommissioning... separative work unit, the common measure by which uranium enrichment services are sold. TESS means the Toll...

METHOD OF PREPARING A FUEL ELEMENT FOR A NUCLEAR REACTOR

DOEpatents

Handwerk, J.H.; BAch, R.A.

1959-08-18

A method is described for preparing a reactor fuel element by forming a mixture of thorium dioxide and an oxide of uranium, the uranium being present. In an oxidation state at least as high as it is in U/sub 3/O/sub 8/, into a desired shape and firing in air at a temperature siifficiently high to reduce the higher uranium oxide to uranium dioxide.

Uranium reduction and resistance to reoxidation under iron-reducing and sulfate-reducing conditions.

PubMed

Boonchayaanant, Benjaporn; Nayak, Dipti; Du, Xin; Criddle, Craig S

2009-10-01

Oxidation and mobilization of microbially-generated U(IV) is of great concern for in situ uranium bioremediation. This study investigated the reoxidation of uranium by oxygen and nitrate in a sulfate-reducing enrichment and an iron-reducing enrichment derived from sediment and groundwater from the Field Research Center in Oak Ridge, Tennessee. Both enrichments were capable of reducing U(VI) rapidly. 16S rRNA gene clone libraries of the two enrichments revealed that Desulfovibrio spp. are dominant in the sulfate-reducing enrichment, and Clostridium spp. are dominant in the iron-reducing enrichment. In both the sulfate-reducing enrichment and the iron-reducing enrichment, oxygen reoxidized the previously reduced uranium but to a lesser extent in the iron-reducing enrichment. Moreover, in the iron-reducing enrichment, the reoxidized U(VI) was eventually re-reduced to its previous level. In both, the sulfate-reducing enrichment and the iron-reducing enrichment, uranium reoxidation did not occur in the presence of nitrate. The results indicate that the Clostridium-dominated iron-reducing communities created conditions that were more favorable for uranium stability with respect to reoxidation despite the fact that fewer electron equivalents were added to these systems. The likely reason is that more of the added electrons are present in a form that can reduce oxygen to water and U(VI) back to U(IV).

Mixed uranium dicarbide and uranium dioxide microspheres and process of making same

DOEpatents

Stinton, David P.

1983-01-01

Nuclear fuel microspheres are made by sintering microspheres containing uranium dioxide and uncombined carbon in a 1 mole percent carbon monoxide/99 mole percent argon atmosphere at 1550.degree. C. and then sintering the microspheres in a 3 mole percent carbon monoxide/97 mole percent argon atmosphere at the same temperature.

PREPARATION OF REFRACTORY OXIDE CRYSTALS

DOEpatents

Grimes, W.R.; Shaffer, J.H.; Watson, G.M.

1962-11-13

A method is given for preparing uranium dioxide, thorium oxide, and beryllium oxide in the form of enlarged individual crystals. The surface of a fused alkali metal halide melt containing dissolved uranium, thorium, or beryllium values is contacted with a water-vapor-bearing inert gas stream at a rate of 5 to 10 cubic centimeters per minute per square centimeter of melt surface area. Growth of individual crystals is obtained by prolonged contact. Beryllium oxide-coated uranium dioxide crystals are prepared by disposing uranium dioxide crystals 5 to 20 microns in diameter in a beryllium-containing melt and contacting the melt with a water-vapor-bearing inert gas stream in the same manner. (AEC)

Influence of uncertainties of isotopic composition of the reprocessed uranium on effectiveness of its enrichment in gas centrifuge cascades

NASA Astrophysics Data System (ADS)

Smirnov, A. Yu; Mustafin, A. R.; Nevinitsa, V. A.; Sulaberidze, G. A.; Dudnikov, A. A.; Gusev, V. E.

2017-01-01

The effect of the uncertainties of the isotopic composition of the reprocessed uranium on its enrichment process in gas centrifuge cascades while diluting it by adding low-enriched uranium (LEU) and waste uranium. It is shown that changing the content of 232U and 236U isotopes in the initial reprocessed uranium within 15% (rel.) can significantly change natural uranium consumption and separative work (up to 2-3%). However, even in case of increase of these parameters is possible to find the ratio of diluents, where the cascade with three feed flows (depleted uranium, LEU and reprocessed uranium) will be more effective than ordinary separation cascade with one feed point for producing LEU from natural uranium.

Molybdenum isotope fractionation during acid leaching of a granitic uranium ore

NASA Astrophysics Data System (ADS)

Migeon, Valérie; Bourdon, Bernard; Pili, Eric; Fitoussi, Caroline

2018-06-01

As an attempt to prevent illicit trafficking of nuclear materials, it is critical to identify the origin and transformation of uranium materials from the nuclear fuel cycle based on chemical and isotope tracers. The potential of molybdenum (Mo) isotopes as tracers is considered in this study. We focused on leaching, the first industrial process used to release uranium from ores, which is also known to extract Mo depending on chemical conditions. Batch experiments were performed in the laboratory with pH ranging from 0.3 to 5.5 in sulfuric acid. In order to span a large range in uranium and molybdenum yields, oxidizers such as nitric acid, hydrogen peroxide and manganese dioxide were also added. An enrichment in heavy Mo isotopes is produced in the solution during leaching of a granitic uranium ore, when Mo recovery is not quantitative. At least two Mo reservoirs were identified in the ore: ∼40% as Mo oxides soluble in water or sulfuric acid, and ∼40% of Mo hosted in sulfides soluble in nitric acid or hydrogen peroxide. At pH > 1.8, adsorption and/or precipitation processes induce a decrease in Mo yields with time correlated with large Mo isotope fractionations. Quantitative models were used to evaluate the relative importance of the processes involved in Mo isotope fractionation: dissolution, adsorption, desorption, precipitation, polymerization and depolymerization. Model best fits are obtained when combining the effects of dissolution/precipitation, and adsorption/desorption onto secondary minerals. These processes are inferred to produce an equilibrium isotope fractionation, with an enrichment in heavy Mo isotopes in the liquid phase and in light isotopes in the solid phase. Quantification of Mo isotope fractionation resulting from uranium leaching is thus a promising tool to trace the origin and transformation of nuclear materials. Our observations of Mo leaching are also consistent with observations of natural Mo isotope fractionation taking place during chemical weathering in terrestrial environments where the role of secondary processes such as adsorption is significant.

77 FR 60482 - Regulatory Guide 5.67, Material Control and Accounting for Uranium Enrichment Facilities...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-10-03

... Accounting for Uranium Enrichment Facilities Authorized To Produce Special Nuclear Material of Low Strategic... Accounting for Uranium Enrichment Facilities Authorized to Produce Special Nuclear Material of Low Strategic... INFORMATION CONTACT: Glenn Tuttle, Office of Nuclear Material Safety and Safeguards, Division of Fuel Cycle...

Mortality (1968-2008) in a French cohort of uranium enrichment workers potentially exposed to rapidly soluble uranium compounds.

PubMed

Zhivin, Sergey; Guseva Canu, Irina; Samson, Eric; Laurent, Olivier; Grellier, James; Collomb, Philippe; Zablotska, Lydia B; Laurier, Dominique

2016-03-01

Until recently, enrichment of uranium for civil and military purposes in France was carried out by gaseous diffusion using rapidly soluble uranium compounds. We analysed the relationship between exposure to soluble uranium compounds and exposure to external γ-radiation and mortality in a cohort of 4688 French uranium enrichment workers who were employed between 1964 and 2006. Data on individual annual exposure to radiological and non-radiological hazards were collected for workers of the AREVA NC, CEA and Eurodif uranium enrichment plants from job-exposure matrixes and external dosimetry records, differentiating between natural, enriched and depleted uranium. Cause-specific mortality was compared with the French general population via standardised mortality ratios (SMR), and was analysed via Poisson regression using log-linear and linear excess relative risk models. Over the period of follow-up, 131 161 person-years at risk were accrued and 21% of the subjects had died. A strong healthy worker effect was observed: all causes SMR=0.69, 95% CI 0.65 to 0.74. SMR for pleural cancer was significantly increased (2.3, 95% CI 1.06 to 4.4), but was only based on nine cases. Internal uranium and external γ-radiation exposures were not significantly associated with any cause of mortality. This is the first study of French uranium enrichment workers. Although limited in statistical power, further follow-up of this cohort, estimation of internal uranium doses and pooling with similar cohorts should elucidate potential risks associated with exposure to soluble uranium compounds. Published by the BMJ Publishing Group Limited. For permission to use (where not already granted under a licence) please go to http://www.bmj.com/company/products-services/rights-and-licensing/

10 CFR 40.33 - Issuance of a license for a uranium enrichment facility.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Issuance of a license for a uranium enrichment facility. 40.33 Section 40.33 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF SOURCE MATERIAL License Applications § 40.33 Issuance of a license for a uranium enrichment facility. (a) The Commission...

78 FR 19311 - Low Enriched Uranium From France; Notice of Commission Determination to Conduct a Full Five-Year...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-03-29

... INTERNATIONAL TRADE COMMISSION [Investigation No. 731-TA-909 (Second Review)] Low Enriched Uranium From France; Notice of Commission Determination to Conduct a Full Five-Year Review AGENCY: United...(c)(5)) to determine whether revocation of the antidumping duty order on low enriched uranium from...

The Military Significance of Small Uranium Enrichment Facilities Fed with Low-Enrichment Uranium (Redacted)

DTIC Science & Technology

1969-12-01

a five-year supply of enriched uranium for reactor fuel . Nevertheless, it seems clear that some foreign enrichment developments are approaching a...produc- tion of fissile material could powerfully influence the assessment of risks and benefits of a nuclear weapons development program . Since... program is likely to include the production of its own relatively pure fissile plutonium. This would involve more rapid cycling and reprocessing of fuel

U.S.-Australia Civilian Nuclear Cooperation: Issues for Congress

DTIC Science & Technology

2010-09-30

7 Uranium Mining and Milling ................................................................................................8...cycle begins with mining uranium ore and upgrading it to yellowcake. Because naturally occurring uranium lacks sufficient fissile 235U to make fuel for...enrichment, and finally fabrication into fuel elements. Australia exports its uranium after the mining and milling stage. Commercial enrichment services

Enriched uranium imports into the EEC countries in 1972 (in German)

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1973-11-01

Parts of a survey published by the statistical Office of the European Communities, entitled ''The Supply of the ECC Countries with Enriched Uranium'' are given and briefly commented on. The main daia and figures on the final utilization of the enriched uranium imported by the EEC countries in 1972 are shown in tabular form. (GE)

75 FR 7525 - Application for a License To Export High-Enriched Uranium

Federal Register 2010, 2011, 2012, 2013, 2014

2010-02-19

... NUCLEAR REGULATORY COMMISSION Application for a License To Export High-Enriched Uranium Pursuant to 10 CFR 110.70(c) ``Public notice of receipt of an application,'' please take notice that the..., February 2, Uranium (93.35%). uranium (87.3 elements in 2010, February 2, 2010, kilograms U-235). France...

Bottling Proliferation of the Uranium Genie: Identifying and Monitoring Clandestine Enrichment Programs

DTIC Science & Technology

2007-04-01

Separation The first method used to enrich uranium on a significant scale was developed by the United States as part of the Manhattan Project during...there does not seem to be a easy way to enrich uranium. It has been over 60 years since the 33 Manhattan Project successfully enriched U-235 to...Proliferation, 91-3. 14 The cost of $5B dollars is adjusted to FY96 dollars. Brookings Institution, “The Costs of the Manhattan Project ,” Global Politics

United States-Gulf Cooperation Council Security Cooperation in a Multipolar World

DTIC Science & Technology

2014-10-01

including plu- tonium separation experiments, uranium enrichment and conversion experiments, and importing various uranium compounds.28 Subsequent...against political protest, a status shared with the two other remaining Arab monarchies, Morocco and Jordan . Geopolitically, the GCC as a region has...commitments, the UAE will not enrich uranium itself, relying instead on imported, enriched fuel. “Abu Dhabi Moves Ahead With Nuclear Program,” Middle

Uranium dioxide electrolysis

DOEpatents

Willit, James L [Batavia, IL; Ackerman, John P [Prescott, AZ; Williamson, Mark A [Naperville, IL

2009-12-29

This is a single stage process for treating spent nuclear fuel from light water reactors. The spent nuclear fuel, uranium oxide, UO.sub.2, is added to a solution of UCl.sub.4 dissolved in molten LiCl. A carbon anode and a metallic cathode is positioned in the molten salt bath. A power source is connected to the electrodes and a voltage greater than or equal to 1.3 volts is applied to the bath. At the anode, the carbon is oxidized to form carbon dioxide and uranium chloride. At the cathode, uranium is electroplated. The uranium chloride at the cathode reacts with more uranium oxide to continue the reaction. The process may also be used with other transuranic oxides and rare earth metal oxides.

SINTERING METAL OXIDES

DOEpatents

Roake, W.E.

1960-09-13

A process is given for producing uranium dioxide material of great density by preparing a compacted mixture of uranium dioxide and from 1 to 3 wt.% of calcium hydride, heating the mixture to at least 675 deg C for decomposition of the hydride and then for sintering, preferably in a vacuum, at from 1550 to 2000 deg C. Calcium metal is formed, some uranium is reduced by the calcium to the metal and a product of high density is obtained.

76 FR 9054 - Notice of Availability of Final Environmental Impact Statement for the AREVA Enrichment Services...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-02-16

... as supplemental information on a proposed electrical transmission line required to power the proposed... proposed uranium enrichment facility. Specifically, AES proposes to use gas centrifuge technology to enrich...; and (3) alternative technologies for uranium enrichment. These alternatives were eliminated from...

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

Federal Register 2010, 2011, 2012, 2013, 2014

2010-07-29

... Uranium Enrichment Fuel Cycle Facility Inspection Reports Regarding Louisiana Energy Services, National... Enrichment Branch, Division of Fuel Cycle Safety and Safeguards, Office of Nuclear Material Safety and... Enrichment Branch, Division of Fuel Cycle Safety and Safeguards, Office of Nuclear Material Safety and...

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

J. L. Dolan; M. J. Marcath; M. Flaska

2014-02-01

A detection system was designed with MCNPX-PoliMi to measure induced-fission neutrons from U-235 and U-238 using active interrogation. Measurements were then performed with this system at the Joint Research Centre (JRC) in Ispra, Italy on low-enriched uranium samples. Liquid scintillators measured induced fission neutron to characterize the samples in terms of their uranium mass and enrichment. Results are presented to investigate and support the use of organic liquid scintillators with active interrogation techniques to characterize uranium containing materials.

4. VIEW OF ROOM 103 IN 1980. SIX OF THE ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

4. VIEW OF ROOM 103 IN 1980. SIX OF THE NINE URANIUM NITRATE STORAGE TANKS ARE SHOWN. HIGHLY ENRICHED URANIUM WAS INTRODUCED INTO THE BUILDING IN THE SUMMER OF 1965 AND THE FIRST EXPERIMENTS WERE PERFORMED IN SEPTEMBER OF 1965. EXPERIMENTS WERE PERFORMED ON ENRICHED URANIUM METAL AND SOLUTION, PLUTONIUM METAL, LOW ENRICHED URANIUM OXIDE, AND SEVERAL SPECIAL APPLICATIONS. AFTER 1983, EXPERIMENTS WERE CONDUCTED PRIMARILY WITH URANYL NITRATE SOLUTIONS, AND DID NOT INVOLVE SOLID MATERIALS. - Rocky Flats Plant, Critical Mass Laboratory, Intersection of Central Avenue & 86 Drive, Golden, Jefferson County, CO

FISSION PRODUCT REMOVAL FROM ORGANIC SOLUTIONS

DOEpatents

Moore, R.H.

1960-05-10

The decontamination of organic solvents from fission products and in particular the treatment of solvents that were used for the extraction of uranium and/or plutonium from aqueous acid solutions of neutron-irradiated uranium are treated. The process broadly comprises heating manganese carbonate in air to a temperature of between 300 and 500 deg C whereby manganese dioxide is formed; mixing the manganese dioxide with the fission product-containing organic solvent to be treated whereby the fission products are precipitated on the manganese dioxide; and separating the fission product-containing manganese dioxide from the solvent.

Thermal properties of nonstoichiometry uranium dioxide

NASA Astrophysics Data System (ADS)

Kavazauri, R.; Pokrovskiy, S. A.; Baranov, V. G.; Tenishev, A. V.

2016-04-01

In this paper, was developed a method of oxidation pure uranium dioxide to a predetermined deviation from the stoichiometry. Oxidation was carried out using the thermogravimetric method on NETZSCH STA 409 CD with a solid electrolyte galvanic cell for controlling the oxygen potential of the environment. 4 samples uranium oxide were obtained with a different ratio of oxygen-to-metal: O / U = 2.002, O / U = 2.005, O / U = 2.015, O / U = 2.033. For the obtained samples were determined basic thermal characteristics of the heat capacity, thermal diffusivity, thermal conductivity. The error of heat capacity determination is equal to 5%. Thermal diffusivity and thermal conductivity of the samples decreased with increasing deviation from stoichiometry. For the sample with O / M = 2.033, difference of both values with those of stoichiometric uranium dioxide is close to 50%.

Challenges dealing with depleted uranium in Germany - Reuse or disposal

DOE Office of Scientific and Technical Information (OSTI.GOV)

Moeller, Kai D.

2007-07-01

During enrichment large amounts of depleted Uranium are produced. In Germany every year 2.800 tons of depleted uranium are generated. In Germany depleted uranium is not classified as radioactive waste but a resource for further enrichment. Therefore since 1996 depleted Uranium is sent to ROSATOM in Russia. However it still has to be dealt with the second generation of depleted Uranium. To evaluate the alternative actions in case a solution has to be found in Germany, several studies have been initiated by the Federal Ministry of the Environment. The work that has been carried out evaluated various possibilities to dealmore » with depleted uranium. The international studies on this field and the situation in Germany have been analyzed. In case no further enrichment is planned the depleted uranium has to be stored. In the enrichment process UF{sub 6} is generated. It is an international consensus that for storage it should be converted to U{sub 3}O{sub 8}. The necessary technique is well established. If the depleted Uranium would have to be characterized as radioactive waste, a final disposal would become necessary. For the planned Konrad repository - a repository for non heat generating radioactive waste - the amount of Uranium is limited by the licensing authority. The existing license would not allow the final disposal of large amounts of depleted Uranium in the Konrad repository. The potential effect on the safety case has not been roughly analyzed. As a result it may be necessary to think about alternatives. Several possibilities for the use of depleted uranium in the industry have been identified. Studies indicate that the properties of Uranium would make it useful in some industrial fields. Nevertheless many practical and legal questions are open. One further option may be the use as shielding e.g. in casks for transport or disposal. Possible techniques for using depleted Uranium as shielding are the use of the metallic Uranium as well as the inclusion in concrete. Another possibility could be the use of depleted uranium for the blending of High enriched Uranium (HEU) or with Plutonium to MOX-elements. (authors)« less

Evaluation of a uranium zirconium hydride fuel rod option for conversion of the MIT research reactor (MITR) from highly-enriched uranium to low-enriched uranium

DOE PAGES

Dunn, F. E.; Wilson, E. H.; Feldman, E. E.; ...

2017-03-23

The conversion of the Massachusetts Institute of Technology Reactor (MITR) from the use of highly-enriched uranium (HEU) fuel-plate assemblies to low-enriched uranium (LEU) by replacing the HEU fuel plates with specially designed General Atomics (GA) uranium zirconium hydride (UZrH) LEU fuel rods is evaluated in this paper. The margin to critical heat flux (CHF) in the core, which is cooled by light water at low pressure, is evaluated analytically for steady-state operation. A form of the Groeneveld CHF lookup table method is used and described in detail. A CHF ratio of 1.41 was found in the present analysis at 10more » MW with engineering hot channel factors included. Therefore, the nominal reactor core power, and neutron flux performance, would need to be reduced by at least 25% in order to meet the regulatory requirement of a minimum CHF ratio of 2.0.« less

Evaluation of a uranium zirconium hydride fuel rod option for conversion of the MIT research reactor (MITR) from highly-enriched uranium to low-enriched uranium

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dunn, F. E.; Wilson, E. H.; Feldman, E. E.

The conversion of the Massachusetts Institute of Technology Reactor (MITR) from the use of highly-enriched uranium (HEU) fuel-plate assemblies to low-enriched uranium (LEU) by replacing the HEU fuel plates with specially designed General Atomics (GA) uranium zirconium hydride (UZrH) LEU fuel rods is evaluated in this paper. The margin to critical heat flux (CHF) in the core, which is cooled by light water at low pressure, is evaluated analytically for steady-state operation. A form of the Groeneveld CHF lookup table method is used and described in detail. A CHF ratio of 1.41 was found in the present analysis at 10more » MW with engineering hot channel factors included. Therefore, the nominal reactor core power, and neutron flux performance, would need to be reduced by at least 25% in order to meet the regulatory requirement of a minimum CHF ratio of 2.0.« less

CONTINUOUS PRECIPITATION METHOD FOR CONVERSION OF URANYL NITRATE TO URANIUM HEXAFLUORIDE

DOEpatents

Reinhart, G.M.; Collopy, T.J.

1962-11-13

A continuous precipitation process is given for converting a uranyl nitrate solution to uranium tetrafluoride. A stream of the uranyl nitrate solution and a stream of an aqueous ammonium hydroxide solution are continuously introduced into an agitated reaction zone maintained at a pH of 5.0 to 6.5. Flow rates are adjusted to provide a mean residence time of the resulting slurry in the reaction zone of at least 30 minutes. After a startup period of two hours the precipitate is recovered from the effluent stream by filtration and is converted to uranium tetrafluoride by reduction to uranium dioxide with hydrogen and reaction of the uranium dioxide with anhydrous hydrogen fluoride. (AEC)

Differential Die-Away Instrument: Report on Fuel Assembly Mock-up Measurements with Neutron Generator

DOE Office of Scientific and Technical Information (OSTI.GOV)

Goodsell, Alison Victoria; Swinhoe, Martyn Thomas; Henzl, Vladimir

2014-09-18

Fresh fuel experiments for the differential die-away (DDA) project were performed using a DT neutron generator, a 15x15 PWR fuel assembly, and nine 3He detectors in a water tank inside of a shielded cell at Los Alamos National Laboratory (LANL). Eight different fuel enrichments were created using low enriched (LEU) and depleted uranium (DU) dioxide fuel rods. A list-mode data acquisition system recorded the time-dependent signal and analysis of the DDA signal die-away time was performed. The die-away time depended on the amount of fissile material in the fuel assembly and the position of the detector. These experiments were performedmore » in support of the spent nuclear fuel Next Generation Safeguards Initiative DDA project. Lessons learned from the fresh fuel DDA instrument experiments and simulations will provide useful information to the spent fuel project.« less

75 FR 9451 - Notice of Receipt and Availability of Environmental Report Supplement 2 for the Proposed GE...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-02

... Availability of Environmental Report Supplement 2 for the Proposed GE-Hitachi Global Laser Enrichment Laser- Based Uranium Enrichment Facility On January 13, 2009, GE-Hitachi Global Laser Enrichment, LLC (GLE) was..., operation, and decommissioning of a laser-based uranium enrichment facility. The proposed facility would be...

Enriched but not depleted uranium affects central nervous system in long-term exposed rat.

PubMed

Houpert, Pascale; Lestaevel, Philippe; Bussy, Cyrill; Paquet, François; Gourmelon, Patrick

2005-12-01

Uranium is well known to induce chemical toxicity in kidneys, but several other target organs, such as central nervous system, could be also affected. Thus in the present study, the effects on sleep-wake cycle and behavior were studied after chronic oral exposure to enriched or depleted uranium. Rats exposed to 4% enriched uranium for 1.5 months through drinking water, accumulated twice as much uranium in some key areas such as the hippocampus, hypothalamus and adrenals than did control rats. This accumulation was correlated with an increase of about 38% of the amount of paradoxical sleep, a reduction of their spatial working memory capacities and an increase in their anxiety. Exposure to depleted uranium for 1.5 months did not induce these effects, suggesting that the radiological activity induces the primary events of these effects of uranium.

Following the electroreduction of uranium dioxide to uranium in LiCl-KCl eutectic in situ using synchrotron radiation

NASA Astrophysics Data System (ADS)

Brown, L. D.; Abdulaziz, R.; Jervis, R.; Bharath, V. J.; Atwood, R. C.; Reinhard, C.; Connor, L. D.; Simons, S. J. R.; Inman, D.; Brett, D. J. L.; Shearing, P. R.

2015-09-01

The electrochemical reduction of uranium dioxide to metallic uranium has been investigated in lithium chloride-potassium chloride eutectic molten salt. Laboratory based electrochemical studies have been coupled with in situ energy dispersive X-ray diffraction, for the first time, to deduce the reduction pathway. No intermediate phases were identified using the X-ray diffraction before, during or after electroreduction to form α-uranium. This suggests that the electrochemical reduction occurs via a single, 4-electron-step, process. The rate of formation of α-uranium is seen to decrease during electrolysis and could be a result of a build-up of oxygen anions in the molten salt. Slow transport of O2- ions away from the UO2 working electrode could impede the electrochemical reduction.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

2008-06-05

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 productmore » 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.« less

U.S.-Australia Civilian Nuclear Cooperation: Issues for Congress

DTIC Science & Technology

2010-12-01

Enrichment.......................................................................................................7 Uranium Mining and Milling...Issues for Congress Congressional Research Service 7 The nuclear fuel cycle begins with mining uranium ore and upgrading it to yellowcake. Because...uranium after the mining and milling stage. Commercial enrichment services are available in the United States, Europe, Russia, and Japan. Fuel

15. DETAILED VIEW OF ENRICHED URANIUM STORAGE TANK. THE ADDITION ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

15. DETAILED VIEW OF ENRICHED URANIUM STORAGE TANK. THE ADDITION OF THE GLASS RINGS SHOWN AT THE TOP OF THE TANK HELPS PREVENT THE URANIUM FROM REACHING CRITICALITY LIMITS. (4/12/62) - Rocky Flats Plant, General Manufacturing, Support, Records-Central Computing, Southern portion of Plant, Golden, Jefferson County, CO

78 FR 75579 - Low Enriched Uranium From France

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-12

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

Theoretical analysis of uranium-doped thorium dioxide: Introduction of a thoria force field with explicit polarization

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shields, A. E.; Ruiz Hernandez, S. E.; Leeuw, N. H. de, E-mail: DeLeeuwN@Cardiff.ac.uk

2015-08-15

Thorium dioxide is used industrially in high temperature applications, but more insight is needed into the behavior of the material as part of a mixed-oxide (MOX) nuclear fuel, incorporating uranium. We have developed a new interatomic potential model including polarizability via a shell model, and commensurate with a prominent existing UO{sub 2} potential, to conduct configurational analyses and to investigate the thermophysical properties of uranium-doped ThO{sub 2}. Using the GULP and Site Occupancy Disorder (SOD) computational codes, we have analyzed the distribution of low concentrations of uranium in the bulk material, where we have not observed the formation of uraniummore » clusters or the dominance of a single preferred configuration. We have calculated thermophysical properties of pure thorium dioxide and Th{sub (1−x)}U{sub x}O{sub 2} which generated values in very good agreement with experimental data.« less

Uranium isotope separation from 1941 to the present

NASA Astrophysics Data System (ADS)

Maier-Komor, Peter

2010-02-01

Uranium isotope separation was the key development for the preparation of highly enriched isotopes in general and thus became the seed for target development and preparation for nuclear and applied physics. In 1941 (year of birth of the author) large-scale development for uranium isotope separation was started after the US authorities were warned that NAZI Germany had started its program for enrichment of uranium and might have confiscated all uranium and uranium mines in their sphere of influence. Within the framework of the Manhattan Projects the first electromagnetic mass separators (Calutrons) were installed and further developed for high throughput. The military aim of the Navy Department was to develop nuclear propulsion for submarines with practically unlimited range. Parallel to this the army worked on the development of the atomic bomb. Also in 1941 plutonium was discovered and the production of 239Pu was included into the atomic bomb program. 235U enrichment starting with natural uranium was performed in two steps with different techniques of mass separation in Oak Ridge. The first step was gas diffusion which was limited to low enrichment. The second step for high enrichment was performed with electromagnetic mass spectrometers (Calutrons). The theory for the much more effective enrichment with centrifugal separation was developed also during the Second World War, but technical problems e.g. development of high speed ball and needle bearings could not be solved before the end of the war. Spying accelerated the development of uranium separation in the Soviet Union, but also later in China, India, Pakistan, Iran and Iraq. In this paper, the physical and chemical procedures are outlined which lead to the success of the project. Some security aspects and Non-Proliferation measures are discussed.

Functionalization of carbon dioxide and carbon disulfide using a stable uranium(III) alkyl complex.

PubMed

Matson, Ellen M; Forrest, William P; Fanwick, Phillip E; Bart, Suzanne C

2011-04-06

A rare uranium(III) alkyl complex, Tp*(2)U(CH(2)Ph) (2) (Tp* = hydrotris(3,5-dimethylpyrazolyl)borate), was synthesized by salt metathesis from Tp*(2)UI (1) and KCH(2)Ph and fully characterized using (1)H NMR, infrared, and electronic absorption spectroscopies as well as X-ray crystallography. This complex has a uranium-carbon distance of 2.57(2) Å, which is comparable to other uranium alkyls reported. Treating this compound with either carbon dioxide or carbon disulfide results in insertion into the uranium-carbon bond to generate Tp*(2)U(κ(2)-O(2)CCH(2)Ph) (3) and Tp*(2)U(SC(S)CH(2)Ph) (4), respectively. These species, characterized spectroscopically and by X-ray crystallography, feature new carboxylate and dithiocarboxylate ligands. Analysis by electronic absorption spectroscopy supports the trivalent oxidation state of the uranium center in both of these derivatives. Addition of trimethylsilylhalides (Me(3)SiX; X = Cl, I) to 3 results in the release of the free silyl ester, Me(3)SiOC(O)CH(2)Ph, forming the initial uranium monohalide species, Tp*(2)UX, which can then be used over multiple cycles for the functionalization of carbon dioxide. © 2011 American Chemical Society

On the reactive occlusion of the (uranium trichloride + lithium chloride + potassium chloride) eutectic salt in zeolite 4A

NASA Astrophysics Data System (ADS)

Lexa, Dusan; Leibowitz, Leonard; Kropf, Jeremy

2000-03-01

The interaction between the (uranium trichloride + lithium chloride + potassium chloride) eutectic salt and zeolite 4A has been studied by temperature-resolved synchrotron powder X-ray diffraction, evolved gas analysis and differential scanning calorimetry, between 300 and 900 K. The onset of salt occlusion by the zeolite has been detected at 450 K. Evidence of a reaction between zeolitic water and uranium trichloride, leading to the formation of uranium dioxide, has appeared at 600 K. The uranium dioxide particle size increases from 2 nm at 600 K to 25 nm at 900 K - an indication of their extra-zeolitic location. No appreciable degradation of the zeolite structure has been observed.

75 FR 60485 - NRC Enforcement Policy Revision

Federal Register 2010, 2011, 2012, 2013, 2014

2010-09-30

... centrifuge or laser enrichment facilities. The NRC has issued licenses for two gas centrifuge uranium... significance)). In addition, the radiological and chemical risks of gas centrifuge uranium enrichment...

Conceptual Core Analysis of Long Life PWR Utilizing Thorium-Uranium Fuel Cycle

NASA Astrophysics Data System (ADS)

Rouf; Su'ud, Zaki

2016-08-01

Conceptual core analysis of long life PWR utilizing thorium-uranium based fuel has conducted. The purpose of this study is to evaluate neutronic behavior of reactor core using combined thorium and enriched uranium fuel. Based on this fuel composition, reactor core have higher conversion ratio rather than conventional fuel which could give longer operation length. This simulation performed using SRAC Code System based on library SRACLIB-JDL32. The calculation carried out for (Th-U)O2 and (Th-U)C fuel with uranium composition 30 - 40% and gadolinium (Gd2O3) as burnable poison 0,0125%. The fuel composition adjusted to obtain burn up length 10 - 15 years under thermal power 600 - 1000 MWt. The key properties such as uranium enrichment, fuel volume fraction, percentage of uranium are evaluated. Core calculation on this study adopted R-Z geometry divided by 3 region, each region have different uranium enrichment. The result show multiplication factor every burn up step for 15 years operation length, power distribution behavior, power peaking factor, and conversion ratio. The optimum core design achieved when thermal power 600 MWt, percentage of uranium 35%, U-235 enrichment 11 - 13%, with 14 years operation length, axial and radial power peaking factor about 1.5 and 1.2 respectively.

Selective Extraction of Uranium from Liquid or Supercritical Carbon Dioxide

DOE Office of Scientific and Technical Information (OSTI.GOV)

Farawila, Anne F.; O'Hara, Matthew J.; Wai, Chien M.

2012-07-31

Current liquid-liquid extraction processes used in recycling irradiated nuclear fuel rely on (1) strong nitric acid to dissolve uranium oxide fuel, and (2) the use of aliphatic hydrocarbons as a diluent in formulating the solvent used to extract uranium. The nitric acid dissolution process is not selective. It dissolves virtually the entire fuel meat which complicates the uranium extraction process. In addition, a solvent washing process is used to remove TBP degradation products, which adds complexity to the recycling plant and increases the overall plant footprint and cost. A liquid or supercritical carbon dioxide (l/sc -CO2) system was designed tomore » mitigate these problems. Indeed, TBP nitric acid complexes are highly soluble in l/sc -CO2 and are capable of extracting uranium directly from UO2, UO3 and U3O8 powders. This eliminates the need for total acid dissolution of the irradiated fuel. Furthermore, since CO2 is easily recycled by evaporation at room temperature and pressure, it eliminates the complex solvent washing process. In this report, we demonstrate: (1) A reprocessing scheme starting with the selective extraction of uranium from solid uranium oxides into a TBP-HNO3 loaded Sc-CO2 phase, (2) Back extraction of uranium into an aqueous phase, and (3) Conversion of recovered purified uranium into uranium oxide. The purified uranium product from step 3 can be disposed of as low level waste, or mixed with enriched uranium for use in a reactor for another fuel cycle. After an introduction on the concept and properties of supercritical fluids, we first report the characterization of the different oxides used for this project. Our extraction system and our online monitoring capability using UV-Vis absorbance spectroscopy directly in sc-CO2 is then presented. Next, the uranium extraction efficiencies and kinetics is demonstrated for different oxides and under different physical and chemical conditions: l/sc -CO2 pressure and temperature, TBP/HNO3 complex used, reductant or complexant used for selectivity, and ionic liquids used as supportive media. To complete the extraction and recovery cycle, we then demonstrate uranium back extraction from the TBP loaded sc-CO2 phase into an aqueous phase and the characterization of the uranium complex formed at the end of this process. Another aspect of this project was to limit proliferation risks by either co-extracting uranium and plutonium, or by leaving plutonium behind by selectively extracting uranium. We report that the former is easily achieved, since plutonium is in the tetravalent or hexavalent oxidation state in the oxidizing environment created by the TBP-nitric acid complex, and is therefore co-extracted. The latter is more challenging, as a reductant or complexant to plutonium has to be used to selectively extract uranium. After undertaking experiments on different reducing or complexing systems (e.g., AcetoHydroxamic Acid (AHA), Fe(II), ascorbic acid), oxalic acid was chosen as it can complex tetravalent actinides (Pu, Np, Th) in the aqueous phase while allowing the extraction of hexavalent uranium in the sc-CO2 phase. Finally, we show results using an alternative media to commonly used aqueous phases: ionic liquids. We show the dissolution of uranium in ionic liquids and its extraction using sc-CO2 with and without the presence of AHA. The possible separation of trivalent actinides from uranium is also demonstrated in ionic liquids using neodymium as a surrogate and diglycolamides as the extractant.« less

Inert matrix fuel in dispersion type fuel elements

NASA Astrophysics Data System (ADS)

Savchenko, A. M.; Vatulin, A. V.; Morozov, A. V.; Sirotin, V. L.; Dobrikova, I. V.; Kulakov, G. V.; Ershov, S. A.; Kostomarov, V. P.; Stelyuk, Y. I.

2006-06-01

The advantages of using inert matrix fuel (IMF) as a dispersion fuel in an aluminium alloy matrix are considered, in particular, low temperatures in the fuel centre, achievable high burn-ups, serviceability in transients and an environmentally friendly process of fuel rod fabrication. Two main versions of IMF are under development at A.A. Bochvar Institute, i.e. heterogeneous or isolated distribution of plutonium. The out-of-pile results on IMF loaded with uranium dioxide as plutonium simulator are presented. Fuel elements with uranium dioxide composition fabricated at A.A. Bochvar Institute are currently under MIR tests (RIAR, Dimitrovgrad). The fuel elements reached a burn-up of 88 MW d kg-1 (equivalent to the burn up of the standard uranium dioxide pelletized fuel) without loss of leak-tightness of the cladding. The feasibility of fabricating IMF of these particular types with plutonium dioxide is considered with a view to in-pile irradiation.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1993-07-01

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

Radiation shielding materials and containers incorporating same

DOEpatents

Mirsky, Steven M.; Krill, Stephen J.; Murray, Alexander P.

2005-11-01

An improved radiation shielding material and storage systems for radioactive materials incorporating the same. The PYRolytic Uranium Compound ("PYRUC") shielding material is preferably formed by heat and/or pressure treatment of a precursor material comprising microspheres of a uranium compound, such as uranium dioxide or uranium carbide, and a suitable binder. The PYRUC shielding material provides improved radiation shielding, thermal characteristic, cost and ease of use in comparison with other shielding materials. The shielding material can be used to form containment systems, container vessels, shielding structures, and containment storage areas, all of which can be used to house radioactive waste. The preferred shielding system is in the form of a container for storage, transportation, and disposal of radioactive waste. In addition, improved methods for preparing uranium dioxide and uranium carbide microspheres for use in the radiation shielding materials are also provided.

Radiation Shielding Materials and Containers Incorporating Same

DOEpatents

Mirsky, Steven M.; Krill, Stephen J.; and Murray, Alexander P.

2005-11-01

An improved radiation shielding material and storage systems for radioactive materials incorporating the same. The PYRolytic Uranium Compound (''PYRUC'') shielding material is preferably formed by heat and/or pressure treatment of a precursor material comprising microspheres of a uranium compound, such as uranium dioxide or uranium carbide, and a suitable binder. The PYRUC shielding material provides improved radiation shielding, thermal characteristic, cost and ease of use in comparison with other shielding materials. The shielding material can be used to form containment systems, container vessels, shielding structures, and containment storage areas, all of which can be used to house radioactive waste. The preferred shielding system is in the form of a container for storage, transportation, and disposal of radioactive waste. In addition, improved methods for preparing uranium dioxide and uranium carbide microspheres for use in the radiation shielding materials are also provided.

DISSOLUTION OF ZIRCONIUM-CONTAINING FUEL ELEMENTS

DOEpatents

Horn, F.L.

1961-12-12

Uranium is recovered from spent uranium fuel elements containing or clad with zirconium. These fuel elements are placed in an anhydrous solution of hydrogen fluoride and nitrogen dioxide. Within this system uranium forms a soluble complex and zirconium forms an insoluble complex. The uranium can then be separated, treated, and removed from solution as uranium hexafluoride. (AEC)

Preliminary study on weapon grade uranium utilization in molten salt reactor miniFUJI

NASA Astrophysics Data System (ADS)

Aji, Indarta Kuncoro; Waris, A.

2014-09-01

Preliminary study on weapon grade uranium utilization in 25MWth and 50MWth of miniFUJI MSR (molten salt reactor) has been carried out. In this study, a very high enriched uranium that we called weapon grade uranium has been employed in UF4 composition. The 235U enrichment is 90 - 95 %. The results show that the 25MWth miniFUJI MSR can get its criticality condition for 1.56 %, 1.76%, and 1.96% of UF4 with 235U enrichment of at least 93%, 90%, and 90%, respectively. In contrast, the 50 MWth miniFUJI reactor can be critical for 1.96% of UF4 with 235U enrichment of at smallest amount 95%. The neutron spectra are almost similar for each power output.

Molecular dynamics analysis of diffusion of uranium and oxygen ions in uranium dioxide

NASA Astrophysics Data System (ADS)

Arima, T.; Yoshida, K.; Idemitsu, K.; Inagaki, Y.; Sato, I.

2010-03-01

Diffusion behaviours of oxygen and uranium were evaluated for bulk and grain-boundaries of uranium dioxide using the molecular dynamics (MD) simulation. It elucidated that oxygen behaved like liquid in superionic state at high temperatures and migrated on sub-lattice sites accompanying formation of lattice defects such as Frenkel defects at middle temperatures. Formation energies of Frenkel and Shottky defects were compared to literature data, and migration energies of oxygen and uranium were estimated by introducing vacancies into the supercell. For grain-boundaries (GB) modelled by the coincidence-site lattice theory, MD calculations showed that GB energy and diffusivities of oxygen and uranium increased with the misorientation angle. By analysing GB structures such as pair-correlation functions, it also showed that the disordered phase was observed for uranium as well as oxygen in GBs especially for a large misorientation angle such as S5 GB. Hence, GB diffusion was much larger than bulk diffusion for oxygen and uranium.

Regularities of spatial association of major endogenous uranium deposits and kimberlitic dykes in the uranium ore regions of the Ukrainian Shield

NASA Astrophysics Data System (ADS)

Kalashnyk, Anna

2015-04-01

During exploration works we discovered the spatial association and proximity time formation of kimberlite dykes (ages are 1,815 and 1,900 Ga for phlogopite) and major industrial uranium deposits in carbonate-sodium metasomatites (age of the main uranium ore of an albititic formation is 1,85-1,70 Ga according to U-Pb method) in Kirovogradsky, Krivorozhsky and Alekseevsko-Lysogorskiy uranium ore regions of the Ukrainian Shield (UkrSh) [1]. In kimberlites of Kirovogradsky ore region uranium content reaches 18-20 g/t. Carbon dioxide is a major component in the formation of hydrothermal uranium deposits and the formation of the sodium in the process of generating the spectrum of alkaline ultrabasic magmas in the range from picritic to kimberlite and this is the connection between these disparate geochemical processes. For industrial uranium deposits in carbonate-sodium metasomatitics of the Kirovogradsky and Krivorozhsky uranium ore regions are characteristic of uranyl carbonate introduction of uranium, which causes correlation between CO2 content and U in range of "poor - ordinary - rich" uranium ore. In productive areas of uranium-ore fields of the Kirovogradsky ore region for phlogopite-carbonate veinlets of uranium ore albitites deep δ13C values (from -7.9 to -6.9o/oo) are characteristic. Isotope-geochemical investigation of albitites from Novokonstantynovskoe, Dokuchaevskoe, Partyzanskoe uranium deposits allowed obtaining direct evidence of the involvement of mantle material during formation of uranium albitites in Kirovogradsky ore region [2]. Petrological characteristics of kimberlites from uranium ore regions of the UkrSh (presence of nodules of dunite and harzburgite garnet in kimberlites, diamonds of peridotite paragenesis, chemical composition of indicator minerals of kimberlite, in particular Gruzskoy areas pyropes (Cr2O3 = 6,1-7,1%, MgO = 19,33-20,01%, CaO = 4,14-4,38 %, the content of knorringite component of most grains > 50mol%), chromites (Cr2O3 = 45,32-62,17%, MgO = 7,3-12,5%) allow us to estimate the depth of generation of kimberlite magmas more than 170-200 km. Ilmenites show two groups according to MgO, Cr2O3 and TiO2 content. Reconstructions of the mantle sections show also two intervals of pressures divided at 4.5 GPa, the upper part is highly metasomatized This high degree metasomatism is determined for almost all mantle columns. It is suggested that large-scale of uranium-bearing mantle fluids may be associated with the ancient degasation during the subduction which is highly enriched in U component . Analysis of the reasons for the marked association kimberlitic dykes and major industrial uranium deposits in carbonate-sodium metasomatic in the UkrSh led to the conclusion that hydrothermal uranium deposits are confined to the supply mantle fluid systems of mantle fault zones exercising brings sodium carbonate solutions enriched uranium from mantle sources. References: 1. Kalashnik A.A. New prognostic-evaluation criteria in technology prognosis of forming industrial endogenous uranium deposits of the Ukrainian Shield, 2014. Scientific proceedings of UkrSGRI, № 2, p. 27-54 (in Russian) 2. Stepanjuk L.M., Bondarenko S.V., Somka V.O. and other, 2012. Source of uranium and uranium-bearing sodium albitites for example of Dokuchaievskogo field of the Ingulsky megablock of the UkrSh: Abstracts of scientific conference "Theoretical issues and research practice metasomatic rocks and ores" (Kyiv, 14-16 March 2012), IGMOF, p.78-80. (in Ukrainian)

METHOD OF FORMING A FUEL ELEMENT FOR A NUCLEAR REACTOR

DOEpatents

Layer, E.H. Jr.; Peet, C.S.

1962-01-23

A method is given for preparing a fuel element for a nuclear reactor. The method includes the steps of sandblasting a body of uranium dioxide to roughen the surface thereof, depositing a thin layer of carbon thereon by thermal decomposition of methane, and cladding the uranium dioxide body with zirconium by gas pressure bonding. (AEC)

Onset conditions for flash sintering of UO 2

DOE PAGES

Raftery, Alicia M.; Pereira da Silva, João Gustavo; Byler, Darrin D.; ...

2017-06-22

In this paper, flash sintering was demonstrated on stoichiometric and non-stoichiometric uranium dioxide pellets at temperatures ranging from room temperature (26°C) up to 600°C. The onset conditions for flash sintering were determined for three stoichiometries (UO 2.00, UO 2.08, and UO 2.16) and analyzed against an established thermal runaway model. The presence of excess oxygen was found to enhance the flash sintering onset behavior of uranium dioxide, lowering the field required to flash and shortening the time required for a flash to occur. Finally, the results from this study highlight the effect of stoichiometry on the flash sintering behavior ofmore » uranium dioxide and will serve as the foundation for future studies on this material.« less

Onset conditions for flash sintering of UO2

NASA Astrophysics Data System (ADS)

Raftery, Alicia M.; Pereira da Silva, João Gustavo; Byler, Darrin D.; Andersson, David A.; Uberuaga, Blas P.; Stanek, Christopher R.; McClellan, Kenneth J.

2017-09-01

In this work, flash sintering was demonstrated on stoichiometric and non-stoichiometric uranium dioxide pellets at temperatures ranging from room temperature (26 °C) up to 600 °C . The onset conditions for flash sintering were determined for three stoichiometries (UO2.00, UO2.08, and UO2.16) and analyzed against an established thermal runaway model. The presence of excess oxygen was found to enhance the flash sintering onset behavior of uranium dioxide, lowering the field required to flash and shortening the time required for a flash to occur. The results from this study highlight the effect of stoichiometry on the flash sintering behavior of uranium dioxide and will serve as the foundation for future studies on this material.

DOE Office of Scientific and Technical Information (OSTI.GOV)

BEHAR, Christophe; GUIBERTEAU, Philippe; DUPERRET, Bernard

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.

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

Federal Register 2010, 2011, 2012, 2013, 2014

2013-04-18

... NUCLEAR REGULATORY COMMISSION [Docket No. 70-3103; NRC-2010-0264] Uranium Enrichment Fuel Cycle Inspection Reports Regarding Louisiana Energy Services, National Enrichment Facility, Eunice, New Mexico..., Division of Fuel Cycle Safety, and Safeguards Office of Nuclear Material Safety, and Safeguards. [FR Doc...

Uranium Conversion & Enrichment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Karpius, Peter Joseph

2017-02-06

The isotopes of uranium that are found in nature, and hence in ‘fresh’ Yellowcake’, are not in relative proportions that are suitable for power or weapons applications. The goal of conversion then is to transform the U 3O 8 yellowcake into UF 6. Conversion and enrichment of uranium is usually required to obtain material with enough 235U to be usable as fuel in a reactor or weapon. The cost, size, and complexity of practical conversion and enrichment facilities aid in nonproliferation by design.

Detection of thermal-induced prompt fission neutrons of highly-enriched uranium: A position sensitive technique

NASA Astrophysics Data System (ADS)

Tartaglione, A.; Di Lorenzo, F.; Mayer, R. E.

2009-07-01

Cargo interrogation in search for special nuclear materials like highly-enriched uranium or 239Pu is a first priority issue of international borders security. In this work we present a thermal-pulsed neutron-based approach to a technique which combines the time-of-flight method and demonstrates a capability to detect small quantities of highly-enriched uranium shielded with high or low Z materials providing, in addition, a manner to know the approximate position of the searched material.

Three-dimensional neutronics optimization of helium-cooled blanket for multi-functional experimental fusion-fission hybrid reactor (FDS-MFX)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jiang, J.; Yuan, B.; Jin, M.

2012-07-01

Three-dimensional neutronics optimization calculations were performed to analyse the parameters of Tritium Breeding Ratio (TBR) and maximum average Power Density (PDmax) in a helium-cooled multi-functional experimental fusion-fission hybrid reactor named FDS (Fusion-Driven hybrid System)-MFX (Multi-Functional experimental) blanket. Three-stage tests will be carried out successively, in which the tritium breeding blanket, uranium-fueled blanket and spent-fuel-fueled blanket will be utilized respectively. In this contribution, the most significant and main goal of the FDS-MFX blanket is to achieve the PDmax of about 100 MW/m3 with self-sustaining tritium (TBR {>=} 1.05) based on the second-stage test with uranium-fueled blanket to check and validate themore » demonstrator reactor blanket relevant technologies based on the viable fusion and fission technologies. Four different enriched uranium materials were taken into account to evaluate PDmax in subcritical blanket: (i) natural uranium, (ii) 3.2% enriched uranium, (iii) 19.75% enriched uranium, and (iv) 64.4% enriched uranium carbide. These calculations and analyses were performed using a home-developed code VisualBUS and Hybrid Evaluated Nuclear Data Library (HENDL). The results showed that the performance of the blanket loaded with 64.4% enriched uranium was the most attractive and it could be promising to effectively obtain tritium self-sufficiency (TBR-1.05) and a high maximum average power density ({approx}100 MW/m{sup 3}) when the blanket was loaded with the mass of {sup 235}U about 1 ton. (authors)« less

75 FR 42466 - Notice of Availability of Draft Environmental Impact Statement and Public Meeting for the AREVA...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-07-21

... electrical transmission line required to power the proposed EREF. On March 17, 2010, the NRC granted an... facility. Specifically, AES proposes to use gas centrifuge technology to enrich the uranium-235 isotope... centrifuge-based technology to enrich the uranium- 235 isotope found in natural uranium to concentrations up...

METHOD FOR PREPARING URANIUM MONOCARBIDE-PLUTONIUM MONOCARBIDE SOLID SOLUTION

DOEpatents

Ogard, A.E.; Leary, J.A.; Maraman, W.J.

1963-03-19

A method is given for preparing solid solutions of uranium monocarbide- plutonium monocarbide. In this method, the powder form of uranium dioxide, plutonium dioxide, and graphite are mixed in a ratio determined by the equation: xUO/sub 2/ + yPuO/sub 2/ + (2+z)C yields UxPu/sub y/C/sub z/ +2CO, where x + y equ al 1.0 and z is greater than 0.9 but less than 1.0. The resulting mixture is compacted and heated in a vacuum at a temperature of 1850 deg C. (AEC)

Raman spectroscopic investigation of thorium dioxide-uranium dioxide (ThO₂-UO₂) fuel materials.

PubMed

Rao, Rekha; Bhagat, R K; Salke, Nilesh P; Kumar, Arun

2014-01-01

Raman spectroscopic investigations were carried out on proposed nuclear fuel thorium dioxide-uranium dioxide (ThO2-UO2) solid solutions and simulated fuels based on ThO2-UO2. Raman spectra of ThO2-UO2 solid solutions exhibited two-mode behavior in the entire composition range. Variations in mode frequencies and relative intensities of Raman modes enabled estimation of composition, defects, and oxygen stoichiometry in these compounds that are essential for their application. The present study shows that Raman spectroscopy is a simple, promising analytical tool for nondestructive characterization of this important class of nuclear fuel materials.

Preliminary study on weapon grade uranium utilization in molten salt reactor miniFUJI

DOE Office of Scientific and Technical Information (OSTI.GOV)

Aji, Indarta Kuncoro; Waris, A., E-mail: awaris@fi.itb.ac.id

Preliminary study on weapon grade uranium utilization in 25MWth and 50MWth of miniFUJI MSR (molten salt reactor) has been carried out. In this study, a very high enriched uranium that we called weapon grade uranium has been employed in UF{sub 4} composition. The {sup 235}U enrichment is 90 - 95 %. The results show that the 25MWth miniFUJI MSR can get its criticality condition for 1.56 %, 1.76%, and 1.96% of UF{sub 4} with {sup 235}U enrichment of at least 93%, 90%, and 90%, respectively. In contrast, the 50 MWth miniFUJI reactor can be critical for 1.96% of UF{sub 4}more » with {sup 235}U enrichment of at smallest amount 95%. The neutron spectra are almost similar for each power output.« less

Investigation of molybdate melts as an alternative method of reprocessing used nuclear fuel

DOE PAGES

Hames, Amber L.; Tkac, Peter; Paulenova, Alena; ...

2017-01-17

Here, an investigation of molybdate melts containing sodium molybdate (Na 2MoO 4) and molybdenum trioxide (MoO 3) to achieve the separation of uranium from fission products by crystallization has been performed. The separation is based on the difference in solubility of the fission product metal oxides compared to the uranium oxide or molybdate in the molybdate melt. The molybdate melt dissolves uranium dioxide at high temperatures, and upon cooling, uranium precipitates as uranium dioxide or molybdate, whereas the fission product metals remain soluble in the melt. Small-scale experiments using gram quantities of uranium dioxide have been performed to investigate themore » feasibility of UO 2 purification from the fission products. The composition of the uranium precipitate as well as data for partitioning of several fission product surrogates between the uranium precipitate and molybdate melt for various melt compositions are presented and discussed. The fission products Cs, Sr, Ru and Rh all displayed very large distribution ratios. The fission products Zr, Pd, and the lanthanides also displayed good distribution ratios (D > 10). A melt consisting of 20 wt% MoO 3-50 wt% Na 2MoO 4-30 wt% UO 2 heated to 1313 K and cooled to 1123 K for the physical separation of the UO 2 product from the melt, and washed once with Na 2MoO 4 displays optimum conditions for separation of the UO 2 from the fission products.« less

Monte Carlo Criticality Analysis of Simple Geometrics COntaining Tungsten Rhenium Alloys Engrained with Uranium Dioxide and Uranium Mononitride

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jonathan A. Webb; Indrajit Charit

2011-08-01

The critical mass and dimensions of simple geometries containing highly enriched uraniumdioxide (UO2) and uraniummononitride (UN) encapsulated in tungsten-rhenium alloys are determined using MCNP5 criticality calculations. Spheres as well as cylinders with length to radius ratios of 1.82 are computationally built to consist of 60 vol.% fuel and 40 vol.% metal matrix. Within the geometries the uranium is enriched to 93 wt.% uranium-235 and the rhenium content within the metal alloy was modeled over a range of 0 to 30 at.%. The spheres containing UO2 were determined to have a critical radius of 18.29 cm to 19.11 cm and amore » critical mass ranging from 366 kg to 424 kg. The cylinders containing UO2 were found to have a critical radius ranging from 17.07 cm to 17.844 cm with a corresponding critical mass of 406 kg to 471 kg. Spheres engrained with UN were determined to have a critical radius ranging from 14.82 cm to 15.19 cm and a critical mass between 222 kg and 242 kg. Cylinders which were engrained with UN were determined to have a critical radius ranging from 13.811 cm to 14.155 cm with a corresponding critical mass of 245 kg to 267 kg. The critical geometries were also computationally submerged in a neutronaically infinite medium of fresh water to determine the effects of rhenium addition on criticality accidents due to water submersion. The monte carlo analysis demonstrated that rhenium addition of up to 30 at.% can reduce the excess reactivity due to water submersion by up to $5.07 for UO2 fueled cylinders, $3.87 for UO2 fueled spheres and approximately $3.00 for UN fueled spheres and cylinders.« less

Neutron source

DOEpatents

Cason, J.L. Jr.; Shaw, C.B.

1975-10-21

A neutron source which is particularly useful for neutron radiography consists of a vessel containing a moderating media of relatively low moderating ratio, a flux trap including a moderating media of relatively high moderating ratio at the center of the vessel, a shell of depleted uranium dioxide surrounding the moderating media of relatively high moderating ratio, a plurality of guide tubes each containing a movable source of neutrons surrounding the flux trap, a neutron shield surrounding one part of each guide tube, and at least one collimator extending from the flux trap to the exterior of the neutron source. The shell of depleted uranium dioxide has a window provided with depleted uranium dioxide shutters for each collimator. Reflectors are provided above and below the flux trap and on the guide tubes away from the flux trap.

Multiple recycle of REMIX fuel based on reprocessed uranium and plutonium mixture in thermal reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fedorov, Y.S.; Bibichev, B.A.; Zilberman, B.Y.

2013-07-01

REMIX fuel consumption in WWER-1000 is considered. REMIX fuel is fabricated from non-separated mixture of uranium and plutonium obtained during NPP spent fuel reprocessing with further makeup by enriched natural uranium. It makes possible to recycle several times the total amount of uranium and plutonium obtained from spent fuel with 100% loading of the WWER-1000 core. The stored SNF could be also involved in REMIX fuel cycle by enrichment of regenerated uranium. The same approach could be applied to closing the fuel cycle of CANDU reactors. (authors)

77 FR 48555 - Agency Information Collection Activities: Proposed Collection; Comment Request

Federal Register 2010, 2011, 2012, 2013, 2014

2012-08-14

... uranium enrichment facility in accordance with 10 CFR Parts 40 and 70. 5. The number of annual respondents... Act of 1954, as amended, and (b) the liability insurance required of uranium enrichment facility...

31 CFR 540.317 - Uranium feed; natural uranium feed.

Code of Federal Regulations, 2013 CFR

2013-07-01

... 31 Money and Finance:Treasury 3 2013-07-01 2013-07-01 false Uranium feed; natural uranium feed... (Continued) OFFICE OF FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.317 Uranium feed; natural uranium feed. The...

31 CFR 540.317 - Uranium feed; natural uranium feed.

Code of Federal Regulations, 2012 CFR

2012-07-01

... 31 Money and Finance:Treasury 3 2012-07-01 2012-07-01 false Uranium feed; natural uranium feed... (Continued) OFFICE OF FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.317 Uranium feed; natural uranium feed. The...

31 CFR 540.317 - Uranium feed; natural uranium feed.

Code of Federal Regulations, 2014 CFR

2014-07-01

... 31 Money and Finance:Treasury 3 2014-07-01 2014-07-01 false Uranium feed; natural uranium feed... (Continued) OFFICE OF FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.317 Uranium feed; natural uranium feed. The...

Experimental study on the measurement of uranium casting enrichment by time-dependent coincidence method

NASA Astrophysics Data System (ADS)

Xie, Wen-Xiong; Li, Jian-Sheng; Gong, Jian; Zhu, Jian-Yu; Huang, Po

2013-10-01

Based on the time-dependent coincidence method, a preliminary experiment has been performed on uranium metal castings with similar quality (about 8-10 kg) and shape (hemispherical shell) in different enrichments using neutron from Cf fast fission chamber and timing DT accelerator. Groups of related parameters can be obtained by analyzing the features of time-dependent coincidence counts between source-detector and two detectors to characterize the fission signal. These parameters have high sensitivity to the enrichment, the sensitivity coefficient (defined as (ΔR/Δm)/R¯) can reach 19.3% per kg of 235U. We can distinguish uranium castings with different enrichments to hold nuclear weapon verification.

10 CFR Appendix F to Part 110 - Illustrative List of Laser-Based Enrichment Plant Equipment and Components Under NRC Export...

Code of Federal Regulations, 2011 CFR

2011-01-01

... come into direct contact with uranium metal vapor or liquid or with process gas consisting of UF6 or a mixture of UF6 and other gases: (1) Uranium vaporization systems (AVLIS). Especially designed or prepared... laser-based enrichment items, the materials resistant to corrosion by the vapor or liquid of uranium...

10 CFR Appendix F to Part 110 - Illustrative List of Laser-Based Enrichment Plant Equipment and Components Under NRC Export...

Code of Federal Regulations, 2014 CFR

2014-01-01

... come into direct contact with uranium metal vapor or liquid or with process gas consisting of UF6 or a mixture of UF6 and other gases: (1) Uranium vaporization systems (AVLIS). Especially designed or prepared... laser-based enrichment items, the materials resistant to corrosion by the vapor or liquid of uranium...

10 CFR Appendix F to Part 110 - Illustrative List of Laser-Based Enrichment Plant Equipment and Components Under NRC Export...

Code of Federal Regulations, 2013 CFR

2013-01-01

... come into direct contact with uranium metal vapor or liquid or with process gas consisting of UF6 or a mixture of UF6 and other gases: (1) Uranium vaporization systems (AVLIS). Especially designed or prepared... laser-based enrichment items, the materials resistant to corrosion by the vapor or liquid of uranium...

10 CFR Appendix F to Part 110 - Illustrative List of Laser-Based Enrichment Plant Equipment and Components Under NRC Export...

Code of Federal Regulations, 2012 CFR

2012-01-01

... come into direct contact with uranium metal vapor or liquid or with process gas consisting of UF6 or a mixture of UF6 and other gases: (1) Uranium vaporization systems (AVLIS). Especially designed or prepared... laser-based enrichment items, the materials resistant to corrosion by the vapor or liquid of uranium...

10 CFR Appendix F to Part 110 - Illustrative List of Laser-Based Enrichment Plant Equipment and Components Under NRC Export...

Code of Federal Regulations, 2010 CFR

2010-01-01

... come into direct contact with uranium metal vapor or liquid or with process gas consisting of UF6 or a mixture of UF6 and other gases: (1) Uranium vaporization systems (AVLIS). Especially designed or prepared... laser-based enrichment items, the materials resistant to corrosion by the vapor or liquid of uranium...

31 CFR 540.309 - Natural uranium.

Code of Federal Regulations, 2013 CFR

2013-07-01

... 31 Money and Finance:Treasury 3 2013-07-01 2013-07-01 false Natural uranium. 540.309 Section 540... FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.309 Natural uranium. The term natural uranium means uranium found in...

31 CFR 540.309 - Natural uranium.

Code of Federal Regulations, 2014 CFR

2014-07-01

... 31 Money and Finance:Treasury 3 2014-07-01 2014-07-01 false Natural uranium. 540.309 Section 540... FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.309 Natural uranium. The term natural uranium means uranium found in...

31 CFR 540.309 - Natural uranium.

Code of Federal Regulations, 2012 CFR

2012-07-01

... 31 Money and Finance:Treasury 3 2012-07-01 2012-07-01 false Natural uranium. 540.309 Section 540... FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.309 Natural uranium. The term natural uranium means uranium found in...

Preliminary investigations on the use of uranium silicide targets for fission Mo-99 production

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cols, H.; Cristini, P.; Marques, R.

1997-08-01

The National Atomic Energy Commission (CNEA) of Argentine Republic owns and operates an installation for production of molybdenum-99 from fission products since 1985, and, since 1991, covers the whole national demand of this nuclide, carrying out a program of weekly productions, achieving an average activity of 13 terabecquerel per week. At present they are finishing an enlargement of the production plant that will allow an increase in the volume of production to about one hundred of terabecquerel. Irradiation targets are uranium/aluminium alloy with 90% enriched uranium with aluminium cladding. In view of international trends held at present for replacing highmore » enrichment uranium (HEU) for enrichment values lower than 20 % (LEU), since 1990 the authors are in contact with the RERTR program, beginning with tests to adapt their separation process to new irradiation target conditions. Uranium silicide (U{sub 3}Si{sub 2}) was chosen as the testing material, because it has an uranium mass per volume unit, so that it allows to reduce enrichment to a value of 20%. CNEA has the technology for manufacturing miniplates of uranium silicide for their purposes. In this way, equivalent amounts of Molybdenum-99 could be obtained with no substantial changes in target parameters and irradiation conditions established for the current process with Al/U alloy. This paper shows results achieved on the use of this new target.« less

3 CFR - Continuation of the National Emergency With Respect to the Disposition of Russian Highly Enriched...

Code of Federal Regulations, 2014 CFR

2014-01-01

... 3 The President 1 2014-01-01 2014-01-01 false Continuation of the National Emergency With Respect to the Disposition of Russian Highly Enriched Uranium Presidential Documents Other Presidential Documents Notice of June 20, 2013 Continuation of the National Emergency With Respect to the Disposition of Russian Highly Enriched Uranium On June 25,...

77 FR 67837 - Agency Information Collection Activities: Submission for the Office of Management and Budget (OMB...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-11-14

... uranium enrichment facility in accordance with 10 CFR Parts 40 and 70. 7. An estimate of the number of... amended, and (b) the liability insurance required of uranium enrichment facility licensees pursuant to...

PRODUCTION OF URANIUM METAL BY CARBON REDUCTION

DOEpatents

Holden, R.B.; Powers, R.M.; Blaber, O.J.

1959-09-22

The preparation of uranium metal by the carbon reduction of an oxide of uranium is described. In a preferred embodiment of the invention a charge composed of carbon and uranium oxide is heated to a solid mass after which it is further heated under vacuum to a temperature of about 2000 deg C to produce a fused uranium metal. Slowly ccoling the fused mass produces a dendritic structure of uranium carbide in uranium metal. Reacting the solidified charge with deionized water hydrolyzes the uranium carbide to finely divide uranium dioxide which can be separated from the coarser uranium metal by ordinary filtration methods.

10 CFR 766.1 - Purpose.

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 4 2011-01-01 2011-01-01 false Purpose. 766.1 Section 766.1 Energy DEPARTMENT OF ENERGY URANIUM ENRICHMENT DECONTAMINATION AND DECOMMISSIONING FUND; PROCEDURES FOR SPECIAL ASSESSMENT OF DOMESTIC... Assessment of domestic utilities for the Uranium Enrichment Decontamination and Decommissioning Fund pursuant...

10 CFR 766.1 - Purpose.

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 4 2014-01-01 2014-01-01 false Purpose. 766.1 Section 766.1 Energy DEPARTMENT OF ENERGY URANIUM ENRICHMENT DECONTAMINATION AND DECOMMISSIONING FUND; PROCEDURES FOR SPECIAL ASSESSMENT OF DOMESTIC... Assessment of domestic utilities for the Uranium Enrichment Decontamination and Decommissioning Fund pursuant...

10 CFR 766.1 - Purpose.

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 4 2012-01-01 2012-01-01 false Purpose. 766.1 Section 766.1 Energy DEPARTMENT OF ENERGY URANIUM ENRICHMENT DECONTAMINATION AND DECOMMISSIONING FUND; PROCEDURES FOR SPECIAL ASSESSMENT OF DOMESTIC... Assessment of domestic utilities for the Uranium Enrichment Decontamination and Decommissioning Fund pursuant...

10 CFR 766.1 - Purpose.

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 4 2013-01-01 2013-01-01 false Purpose. 766.1 Section 766.1 Energy DEPARTMENT OF ENERGY URANIUM ENRICHMENT DECONTAMINATION AND DECOMMISSIONING FUND; PROCEDURES FOR SPECIAL ASSESSMENT OF DOMESTIC... Assessment of domestic utilities for the Uranium Enrichment Decontamination and Decommissioning Fund pursuant...

10 CFR 766.1 - Purpose.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 4 2010-01-01 2010-01-01 false Purpose. 766.1 Section 766.1 Energy DEPARTMENT OF ENERGY URANIUM ENRICHMENT DECONTAMINATION AND DECOMMISSIONING FUND; PROCEDURES FOR SPECIAL ASSESSMENT OF DOMESTIC... Assessment of domestic utilities for the Uranium Enrichment Decontamination and Decommissioning Fund pursuant...

Recent advances in the study of the UO2-PuO2 phase diagram at high temperatures

NASA Astrophysics Data System (ADS)

Böhler, R.; Welland, M. J.; Prieur, D.; Cakir, P.; Vitova, T.; Pruessmann, T.; Pidchenko, I.; Hennig, C.; Guéneau, C.; Konings, R. J. M.; Manara, D.

2014-05-01

Recently, novel container-less laser heating experimental data have been published on the melting behaviour of pure PuO2 and PuO2-rich compositions in the uranium dioxide-plutonium dioxide system. Such data showed that previous data obtained by more traditional furnace heating techniques were affected by extensive interaction between the sample and its containment. It is therefore paramount to check whether data so far used by nuclear engineers for the uranium-rich side of the pseudo-binary dioxide system can be confirmed or not. In the present work, new data are presented both in the UO2-rich part of the phase diagram, most interesting for the uranium-plutonium dioxide based nuclear fuel safety, and in the PuO2 side. The new results confirm earlier furnace heating data in the uranium-dioxide rich part of the phase diagram, and more recent laser-heating data in the plutonium-dioxide side of the system. As a consequence, it is also confirmed that a minimum melting point must exist in the UO2-PuO2 system, at a composition between x(PuO2) = 0.4 and x(PuO2) = 0.7 and 2900 K ⩽ T ⩽ 3000 K. Taking into account that, especially at high temperature, oxygen chemistry has an effect on the reported phase boundary uncertainties, the current results should be projected in the ternary U-Pu-O system. This aspect has been extensively studied here by X-ray diffraction and X-ray absorption spectroscopy. The current results suggest that uncertainty bands related to oxygen behaviour in the equilibria between condensed phases and gas should not significantly affect the qualitative trend of the current solid-liquid phase boundaries.

CHEMICAL DIFFERENCES BETWEEN SLUDGE SOLIDS AT THE F AND H AREA TANK FARMS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Reboul, S.

2012-08-29

The primary source of waste solids received into the F Area Tank Farm (FTF) was from PUREX processing performed to recover uranium and plutonium from irradiated depleted uranium targets. In contrast, two primary sources of waste solids were received into the H Area Tank Farm (HTF): a) waste from PUREX processing; and b) waste from H-modified (HM) processing performed to recover uranium and neptunium from burned enriched uranium fuel. Due to the differences between the irradiated depleted uranium targets and the burned enriched uranium fuel, the average compositions of the F and H Area wastes are markedly different from onemore » another. Both F and H Area wastes contain significant amounts of iron and aluminum compounds. However, because the iron content of PUREX waste is higher than that of HM waste, and the aluminum content of PUREX waste is lower than that of HM waste, the iron to aluminum ratios of typical FTF waste solids are appreciably higher than those of typical HTF waste solids. Other constituents present at significantly higher concentrations in the typical FTF waste solids include uranium, nickel, ruthenium, zinc, silver, cobalt and copper. In contrast, constituents present at significantly higher concentrations in the typical HTF waste solids include mercury, thorium, oxalate, and radionuclides U-233, U-234, U-235, U-236, Pu-238, Pu-242, Cm-244, and Cm-245. Because of the higher concentrations of Pu-238 in HTF, the long-term concentrations of Th-230 and Ra-226 (from Pu-238 decay) will also be higher in HTF. The uranium and plutonium distributions of the average FTF waste were found to be consistent with depleted uranium and weapons grade plutonium, respectively (U-235 comprised 0.3 wt% of the FTF uranium, and Pu-240 comprised 6 wt% of the FTF plutonium). In contrast, at HTF, U-235 comprised 5 wt% of the uranium, and Pu-240 comprised 17 wt% of the plutonium, consistent with enriched uranium and high burn-up plutonium. X-ray diffraction analyses of various FTF and HTF samples indicated that the primary crystalline compounds of iron in sludge solids are Fe{sub 2}O{sub 3}, Fe{sub 3}O{sub 4}, and FeO(OH), and the primary crystalline compounds of aluminum are Al(OH){sub 3} and AlO(OH). Also identified were carbonate compounds of calcium, magnesium, and sodium; a nitrated sodium aluminosilicate; and various uranium compounds. Consistent with expectations, oxalate compounds were identified in solids associated with oxalic acid cleaning operations. The most likely oxidation states and chemical forms of technetium are assessed in the context of solubility, since technetium-99 is a key risk driver from an environmental fate and transport perspective. The primary oxidation state of technetium in SRS sludge solids is expected to be Tc(IV). In salt waste, the primary oxidation state is expected to be Tc(VII). The primary form of technetium in sludge is expected to be a hydrated technetium dioxide, TcO{sub 2} {center_dot} xH{sub 2}O, which is relatively insoluble and likely co-precipitated with iron. In salt waste solutions, the primary form of technetium is expected to be the very soluble pertechnetate anion, TcO{sub 4}{sup -}. The relative differences between the F and H Tank Farm waste provide a basis for anticipating differences that will occur as constituents of FTF and HTF waste residue enter the environment over the long-term future. If a constituent is significantly more dominant in one of the Tank Farms, its long-term environmental contribution will likely be commensurately higher, assuming the environmental transport conditions of the two Tank Farms share some commonality. It is in this vein that the information cited in this document is provided - for use during the generation, assessment, and validation of Performance Assessment modeling results.« less

10 CFR 75.4 - Definitions.

Code of Federal Regulations, 2011 CFR

2011-01-01

... to IAEA Safeguards) means the collection of environmental samples (e.g., air, water, vegetation, soil... uranium or enriching uranium in the isotope 235, zirconium tubes, heavy water or deuterium, nuclear-grade...); (3) A fuel fabrication plant; (4) An enrichment plant or isotope separation plant for the separation...

5. VIEW OF THE FOUNDRY. IN THE FOUNDRY, ENRICHED URANIUM ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

5. VIEW OF THE FOUNDRY. IN THE FOUNDRY, ENRICHED URANIUM WAS CAST INTO SLABS OR INGOTS FROM WHICH WEAPONS COMPONENTS WERE FABRICATED. (4/4/66) - Rocky Flats Plant, General Manufacturing, Support, Records-Central Computing, Southern portion of Plant, Golden, Jefferson County, CO

4. VIEW OF THE FOUNDRY. IN THE FOUNDRY, ENRICHED URANIUM ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

4. VIEW OF THE FOUNDRY. IN THE FOUNDRY, ENRICHED URANIUM WAS CAST INTO SLABS OR INGOTS FROM WHICH WEAPONS COMPONENTS WERE FABRICATED. (5/17/62). - Rocky Flats Plant, General Manufacturing, Support, Records-Central Computing, Southern portion of Plant, Golden, Jefferson County, CO

Nickel container of highly-enriched uranium bodies and sodium

DOEpatents

Zinn, Walter H.

1976-01-01

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.

Iran’s Reemergence as a Major Player in Global Security

DTIC Science & Technology

2013-05-21

economic sanctions levied against the Islamic Republic. Iran continues to deny International Atomic Energy Agency inspectors’ access to possible uranium ...build nuclear weapons.”55 Mr. Clapper went on to say that “Iran’s technical advancement, particularly in uranium enrichment, strengthens our assessment...will to do so.”56 During the briefing, he made clear that Iran is technically capable of producing enough highly enriched uranium for a weapon

31 CFR 540.318 - Uranium Hexafluoride (UF6).

Code of Federal Regulations, 2012 CFR

2012-07-01

... 31 Money and Finance:Treasury 3 2012-07-01 2012-07-01 false Uranium Hexafluoride (UF6). 540.318... OF FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.318 Uranium Hexafluoride (UF6). The term uranium...

31 CFR 540.318 - Uranium Hexafluoride (UF6).

Code of Federal Regulations, 2011 CFR

2011-07-01

... 31 Money and Finance:Treasury 3 2011-07-01 2011-07-01 false Uranium Hexafluoride (UF6). 540.318... OF FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.318 Uranium Hexafluoride (UF6). The term uranium...

31 CFR 540.318 - Uranium Hexafluoride (UF6).

Code of Federal Regulations, 2013 CFR

2013-07-01

... 31 Money and Finance:Treasury 3 2013-07-01 2013-07-01 false Uranium Hexafluoride (UF6). 540.318... OF FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.318 Uranium Hexafluoride (UF6). The term uranium...

31 CFR 540.318 - Uranium Hexafluoride (UF6).

Code of Federal Regulations, 2014 CFR

2014-07-01

... 31 Money and Finance:Treasury 3 2014-07-01 2014-07-01 false Uranium Hexafluoride (UF6). 540.318... OF FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.318 Uranium Hexafluoride (UF6). The term uranium...

31 CFR 540.318 - Uranium Hexafluoride (UF6).

Code of Federal Regulations, 2010 CFR

2010-07-01

... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false Uranium Hexafluoride (UF6). 540.318... OF FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.318 Uranium Hexafluoride (UF6). The term uranium...

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

DOEpatents

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

1995-05-02

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

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

DOEpatents

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

1995-01-01

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

The Proliferation Security Initiative: A Means to an End for the Operational Commander

DTIC Science & Technology

2009-05-04

The Reduced Enrichment for Research and Test Reactors ( RERTR ) Program develops technology necessary to enable the conversion of civilian...facilities using high enriched uranium (HEU) to low enriched uranium (LEU) fuels and targets. The RERTR Program was initiated by the U.S. Department of...processes have been developed for producing radioisotopes with LEU targets. The RERTR Program is managed by the Office of Nuclear Material Threat

Proposal for Monitoring Within the Centrifuge Cascades of Uranium Enrichment Facilities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Farrar, David R.

2017-04-01

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

Hybrid Interferometric/Dispersive Atomic Spectroscopy For Nuclear Materials Analysis

NASA Astrophysics Data System (ADS)

Morgan, Phyllis K.

Laser-induced breakdown spectroscopy (LIBS) is an optical emission spectroscopy technique that holds promise for detection and rapid analysis of elements relevant for nuclear safeguards and nonproliferation, including the measurement of isotope ratios. One important application of LIBS is the measurement of uranium enrichment (235U/238U), which requires high spectral resolution (e.g., 25 pm for the 424.437 nm U II line). Measuring uranium enrichment is important in nuclear nonproliferation and safeguards because the uranium highly enriched in the 235U isotope can be used to construct nuclear weapons. High-resolution dispersive spectrometers necessary for such measurements are typically bulky and expensive. A hybrid interferometric/dispersive spectrometer prototype, which consists of an inexpensive, compact Fabry-Perot etalon integrated with a low to moderate resolution Czerny-Turner spectrometer, was assembled for making high-resolution measurements of nuclear materials in a laboratory setting. To more fully take advantage of this low-cost, compact hybrid spectrometer, a mathematical reconstruction technique was developed to accurately reconstruct relative line strengths from complex spectral patterns with high resolution. Measurement of the mercury 313.1555/313.1844 nm doublet from a mercury-argon lamp yielded a spectral line intensity ratio of 0.682, which agrees well with an independent measurement by an echelle spectrometer and previously reported values. The hybrid instrument was used in LIBS measurements and achieved the resolution needed for isotopic selectivity of LIBS of uranium in ambient air. The samples used were a natural uranium foil (0.7% of 235U) and a uranium foil highly enriched in 235U to 93%. Both samples were provided by the Penn State University's Breazeale Nuclear Reactor. The enrichment of the uranium foils was verified using a high-purity germanium detector and dedicated software for multi-group spectral analysis. Uranium spectral line widths of ˜10 pm were measured at a center wavelength 424.437 nm, clearly discriminating the natural from the highly enriched uranium at that wavelength. The 424.167 nm isotope shift (˜6 pm), limited by spectral broadening, was only partially resolved but still discernible. This instrument and reconstruction method could enable the design of significantly smaller, portable high-resolution instruments with isotopic specificity, benefiting nuclear safeguards, treaty verification, nuclear forensics, and a variety of other spectroscopic applications.

Characterization of uranium surfaces machined with aqueous propylene glycol-borax or perchloroethylene-mineral oil coolants

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cristy, S.S.; Bennett, R.K. Jr.; Dillon, J.J.

1986-12-31

The use of perchloroethylene (perc) as an ingredient in coolants for machining enriched uranium at the Oak Ridge Y-12 Plant has been discontinued because of environmental concerns. A new coolant was substituted in December 1985, which consists of an aqueous solution of propylene glycol with borax (sodium tetraborate) added as a nuclear poison and with a nitrite added as a corrosion inhibitor. Uranium surfaces machined using the two coolants were compared with respects to residual contamination, corrosion or corrosion potential, and with the aqueous propylene glycol-borax coolant was found to be better than that of enriched uranium machined with themore » perc-mineral oil coolant. The boron residues on the final-finished parts machined with the borax-containing coolant were not sufficient to cause problems in further processing. All evidence indicated that the enriched uranium surfaces machined with the borax-containing coolant will be as satisfactory as those machined with the perc coolant.« less

Performance Evaluation of Spectroscopic Detectors for LEU Hold-up Measurements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Venkataraman, Ramkumar; Nutter, Greg; McElroy, Robert Dennis

The hold-up measurement of low-enriched uranium materials may require use of alternate detector types relative to the measurement of highly enriched uranium. This is in part due to the difference in process scale (i.e., the components are generally larger for low-enriched uranium systems), but also because the characteristic gamma-ray lines from 235U used for assay of highly enriched uranium will be present at a much reduced intensity (on a per gram of uranium basis) at lower enrichments. Researchers at Oak Ridge National Laboratory examined the performance of several standard detector types, e.g., NaI(Tl), LaBr3(Ce), and HPGe, to select a suitablemore » candidate for measuring and quantifying low-enriched uranium hold-up in process pipes and equipment at the Portsmouth gaseous diffusion plant. Detector characteristics, such as energy resolution (full width at half maximum) and net peak count rates at gamma ray energies spanning a range of 60–1332 keV, were measured for the above-mentioned detector types using the same sources and in the same geometry. Uranium enrichment standards (Certified Reference Material no. 969 and Certified Reference Material no. 146) were measured using each of the detector candidates in the same geometry. The net count rates recorded by each detector at 186 keV and 1,001 keV were plotted as a function of enrichment (atom percentage). Background measurements were made in unshielded and shielded configurations under both ambient and elevated conditions of 238U activity. The highly enriched uranium hold-up measurement campaign at the Portsmouth plant was performed on process equipment that had been cleaned out. Therefore, in most cases, the thickness of the uranium deposits was less than the “infinite thickness” for the 186 keV gamma rays to be completely self-attenuated. Because of this, in addition to measuring the 186 keV gamma, the 1,001 keV gamma ray from 234mPa—a daughter of 238U in secular equilibrium with its parent—will also need to be measured. Based on the performance criteria of detection efficiency, energy resolution, peak-to-continuum ratios, minimum detectable limits, and the weight of the shielded probe, a shielded (0.5 in. thick lead shield) 2 × 2 in. NaI(Tl) detector is recommended for use. The recommended approach is to carry out analysis using data from both 186 keV and 1,001 keV gamma rays, and select a best result based on propagated uncertainty estimates. It is also highly recommended that a two-point gain stabilization scheme based on an 241Am seed embedded in the probe be implemented. Shielding configurations to reduce the impact of background interference on the measurement of 1,001 keV gamma-ray are discussed.« less

49 CFR 173.417 - Authorized fissile materials packages.

Code of Federal Regulations, 2010 CFR

2010-10-01

... for export and import shipments. (2) A residual “heel” of enriched solid uranium hexafluoride may be... “Heel” in a Specification 7A Cylinder) Maximum cylinder diameter Centimeters Inches Cylinder volume Liters Cubic feet Maximum Uranium 235-enrichment (weight)percent Maximum “Heel” weight per cylinder UF6...

49 CFR 173.417 - Authorized fissile materials packages.

Code of Federal Regulations, 2011 CFR

2011-10-01

... for export and import shipments. (2) A residual “heel” of enriched solid uranium hexafluoride may be... “Heel” in a Specification 7A Cylinder) Maximum cylinder diameter Centimeters Inches Cylinder volume Liters Cubic feet Maximum Uranium 235-enrichment (weight)percent Maximum “Heel” weight per cylinder UF6...

PROCESS OF PRODUCING REFRACTORY URANIUM OXIDE ARTICLES

DOEpatents

Hamilton, N.E.

1957-12-01

A method is presented for fabricating uranium oxide into a shaped refractory article by introducing a uranium halide fluxing reagent into the uranium oxide, and then mixing and compressing the materials into a shaped composite mass. The shaped mass of uranium oxide and uranium halide is then fired at an elevated temperature so as to form a refractory sintered article. It was found in the present invention that the introduction of a uraninm halide fluxing agent afforded a fluxing action with the uranium oxide particles and that excellent cohesion between these oxide particles was obtained. Approximately 90% of uranium dioxide and 10% of uranium tetrafluoride represent a preferred composition.

235U enrichment determination on UF6 cylinders with CZT detectors

NASA Astrophysics Data System (ADS)

Berndt, Reinhard; Mortreau, Patricia

2018-04-01

Measurements of uranium enrichment in UF6 transit cylinders are an important nuclear safeguards verification task, which is performed using a non-destructive assay method, the traditional enrichment meter, which involves measuring the count rate of the 186 keV gamma ray. This provides a direct measure of the 235U enrichment. Measurements are typically performed using either high-resolution detectors (Germanium) with e-cooling and battery operation, or portable devices equipped with low resolution detectors (NaI). Despite good results being achieved when measuring Low Enriched Uranium in 30B type cylinders and natural uranium in 48Y type containers using both detector systems, there are situations, which preclude the use of one or both of these systems. The focus of this work is to address some of the recognized limitations in relation to the current use of the above detector systems by considering the feasibility of an inspection instrument for 235U enrichment measurements on UF6 cylinders using the compact and light Cadmium Zinc Telluride (CZT) detectors. In the present work, test measurements were carried out, under field conditions and on full-size objects, with different CZT detectors, in particular for situations where existing systems cannot be used e.g. for stacks of 48Y type containers with depleted uranium. The main result of this study shows that the CZT detectors, actually a cluster of four μCZT1500 micro spectrometers provide as good results as the germanium detector in the ORTEC Micro-trans SPEC HPGe Portable spectrometer, and most importantly in particular for natural and depleted uranium in 48Y cylinders.

Lattice constant in nonstoichiometric uranium dioxide from first principles

NASA Astrophysics Data System (ADS)

Bruneval, Fabien; Freyss, Michel; Crocombette, Jean-Paul

2018-02-01

Nonstoichiometric uranium dioxide experiences a shrinkage of its lattice constant with increasing oxygen content, in both the hypostoichiometric and the hyperstoichiometric regimes. Based on first-principles calculations within the density functional theory (DFT)+U approximation, we have developed a point defect model that accounts for the volume of relaxation of the most significant intrinsic defects of UO2. Our point defect model takes special care of the treatment of the charged defects in the equilibration of the model and in the determination of reliable defect volumes of formation. In the hypostoichiometric regime, the oxygen vacancies are dominant and explain the lattice constant variation with their surprisingly positive volume of relaxation. In the hyperstoichiometric regime, the uranium vacancies are predicted to be the dominating defect,in contradiction with experimental observations. However, disregarding uranium vacancies allows us to recover a good match for the lattice-constant variation as a function of stoichiometry. This can be considered a clue that the uranium vacancies are indeed absent in UO2 +x, possibly due to the very slow diffusion of uranium.

31 CFR 540.315 - Uranium-235 (U235).

Code of Federal Regulations, 2013 CFR

2013-07-01

... 31 Money and Finance:Treasury 3 2013-07-01 2013-07-01 false Uranium-235 (U235). 540.315 Section... FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.315 Uranium-235 (U235). The term uranium-235 or U235 means the fissile...

31 CFR 540.315 - Uranium-235 (U235).

Code of Federal Regulations, 2012 CFR

2012-07-01

... 31 Money and Finance:Treasury 3 2012-07-01 2012-07-01 false Uranium-235 (U235). 540.315 Section... FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.315 Uranium-235 (U235). The term uranium-235 or U235 means the fissile...

49 CFR 173.434 - Activity-mass relationships for uranium and natural thorium.

Code of Federal Regulations, 2014 CFR

2014-10-01

... 49 Transportation 2 2014-10-01 2014-10-01 false Activity-mass relationships for uranium and....434 Activity-mass relationships for uranium and natural thorium. The table of activity-mass relationships for uranium and natural thorium are as follows: Thorium and uranium enrichment 1(Wt% 235 U present...

49 CFR 173.434 - Activity-mass relationships for uranium and natural thorium.

Code of Federal Regulations, 2012 CFR

2012-10-01

... 49 Transportation 2 2012-10-01 2012-10-01 false Activity-mass relationships for uranium and....434 Activity-mass relationships for uranium and natural thorium. The table of activity-mass relationships for uranium and natural thorium are as follows: Thorium and uranium enrichment 1(Wt% 235 U present...

49 CFR 173.434 - Activity-mass relationships for uranium and natural thorium.

Code of Federal Regulations, 2013 CFR

2013-10-01

... 49 Transportation 2 2013-10-01 2013-10-01 false Activity-mass relationships for uranium and....434 Activity-mass relationships for uranium and natural thorium. The table of activity-mass relationships for uranium and natural thorium are as follows: Thorium and uranium enrichment 1(Wt% 235 U present...

31 CFR 540.315 - Uranium-235 (U235).

Code of Federal Regulations, 2014 CFR

2014-07-01

... 31 Money and Finance:Treasury 3 2014-07-01 2014-07-01 false Uranium-235 (U235). 540.315 Section... FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.315 Uranium-235 (U235). The term uranium-235 or U235 means the fissile...

31 CFR 540.315 - Uranium-235 (U235).

Code of Federal Regulations, 2011 CFR

2011-07-01

... 31 Money and Finance:Treasury 3 2011-07-01 2011-07-01 false Uranium-235 (U235). 540.315 Section... FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.315 Uranium-235 (U235). The term uranium-235 or U235 means the fissile...

31 CFR 540.315 - Uranium-235 (U235).

Code of Federal Regulations, 2010 CFR

2010-07-01

... 31 Money and Finance: Treasury 3 2010-07-01 2010-07-01 false Uranium-235 (U235). 540.315 Section... FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.315 Uranium-235 (U235). The term uranium-235 or U235 means the fissile...

RAND Review: Volume 29, Number 2, Summer 2005

DTIC Science & Technology

2005-01-01

is problematic because al Qaeda "Protecting businesses against tinued reliance on martyrdom; and " franchises " its attacks to local the economic impact...enriching uranium. We’ve got a lot ofnatural answered, "you would fee! safer if you had nuclear uranium. It’s legal. We want to enrich Uranium.’ And weapons...is then safer . If Iran adds nuclear weapons to its civil war within Islam rather than a global war on ter- arsenal, they already have Israel to worry

Three-dimensional ordered titanium dioxide-zirconium dioxide film-based microfluidic device for efficient on-chip phosphopeptide enrichment.

PubMed

Zhao, De; He, Zhongyuan; Wang, Gang; Wang, Hongzhi; Zhang, Qinghong; Li, Yaogang

2016-09-15

Microfluidic technology plays a significant role in separating biomolecules, because of its miniaturization, integration, and automation. Introducing micro/nanostructured functional materials can improve the properties of microfluidic devices, and extend their application. Inverse opal has a three-dimensional ordered net-like structure. It possesses a large surface area and exhibits good mass transport, making it a good candidate for bio-separation. This study exploits inverse opal titanium dioxide-zirconium dioxide films for on-chip phosphopeptide enrichment. Titanium dioxide-zirconium dioxide inverse opal film-based microfluidic devices were constructed from templates of 270-, 340-, and 370-nm-diameter poly(methylmethacrylate) spheres. The phosphopeptide enrichments of these devices were determined by matrix-assisted laser desorption/ionization time-of-flight (MALDI-TOF) mass spectrometry. The device constructed from the 270-nm-diameter sphere template exhibited good comprehensive phosphopeptide enrichment, and was the best among these three devices. Because the size of opal template used in construction was the smallest, the inverse opal film therefore had the smallest pore sizes and the largest surface area. Enrichment by this device was also better than those of similar devices based on nanoparticle films and single component films. The titanium dioxide-zirconium dioxide inverse opal film-based device provides a promising approach for the efficient separation of various biomolecules. Copyright © 2016 Elsevier Inc. All rights reserved.

77 FR 33253 - Regulatory Guide 8.24, Revision 2, Health Physics Surveys During Enriched Uranium-235 Processing...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-06-05

... NUCLEAR REGULATORY COMMISSION [NRC-2010-0115] Regulatory Guide 8.24, Revision 2, Health Physics..., ``Health Physics Surveys During Enriched Uranium-235 Processing and Fuel Fabrication'' was issued with a... specifically with the following aspects of an acceptable occupational health physics program that are closely...

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

This article is a review of the agreement between the United States and two of the former Soviet republics to buy and convert weapons-grade uranium into reactor fuel. Under this 20 year agreement, the US Enrichment Corporation will buy 500 metric tons for a price of $11.9B. This will convert into 15,260 tons of low-enriched uranium.

10 CFR Appendix E to Part 110 - Illustrative List of Chemical Exchange or Ion Exchange Enrichment Plant Equipment and Components...

Code of Federal Regulations, 2010 CFR

2010-01-01

... designed or prepared electrochemical reduction cells to reduce uranium from one valence state to another for uranium enrichment using the chemical exchange process. The cell materials in contact with process solutions must be corrosion resistant to concentrated hydrochloric acid solutions. The cell cathodic...

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

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 2 2013-01-01 2013-01-01 false Nuclear material control and accounting for uranium enrichment facilities authorized to produce special nuclear material of low strategic significance. 74.33 Section 74.33 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) MATERIAL CONTROL AND ACCOUNTING OF SPECIAL...

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

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 2 2014-01-01 2014-01-01 false Nuclear material control and accounting for uranium enrichment facilities authorized to produce special nuclear material of low strategic significance. 74.33 Section 74.33 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) MATERIAL CONTROL AND ACCOUNTING OF SPECIAL...

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

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 2 2012-01-01 2012-01-01 false Nuclear material control and accounting for uranium enrichment facilities authorized to produce special nuclear material of low strategic significance. 74.33 Section 74.33 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) MATERIAL CONTROL AND ACCOUNTING OF SPECIAL...

Protein Hydrogel Microbeads for Selective Uranium Mining from Seawater.

PubMed

Kou, Songzi; Yang, Zhongguang; Sun, Fei

2017-01-25

Practical methods for oceanic uranium extraction have yet to be developed in order to tap into the vast uranium reserve in the ocean as an alternative energy. Here we present a protein hydrogel system containing a network of recently engineered super uranyl binding proteins (SUPs) that is assembled through thiol-maleimide click chemistry under mild conditions. Monodisperse SUP hydrogel microbeads fabricated by a microfluidic device further enable uranyl (UO 2 2+ ) enrichment from natural seawater with great efficiency (enrichment index, K = 2.5 × 10 3 ) and selectivity. Our results demonstrate the feasibility of using protein hydrogels to extract uranium from the ocean.

Two-dimensional fluorescence spectroscopy of uranium isotopes in femtosecond laser ablation plumes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Phillips, Mark C.; Brumfield, Brian E.; LaHaye, Nicole

Here, we demonstrate measurement of uranium isotopes in femtosecond laser ablation plumes using two-dimensional fluorescence spectroscopy (2DFS). The high-resolution, tunable CW-laser spectroscopy technique clearly distinguishes atomic absorption from 235U and 238U in natural and highly enriched uranium metal samples. We present analysis of spectral resolution and analytical performance of 2DFS as a function of ambient pressure. Simultaneous measurement using time-resolved absorption spectroscopy provides information on temporal dynamics of the laser ablation plume and saturation behavior of fluorescence signals. The rapid, non-contact measurement is promising for in-field, standoff measurements of uranium enrichment for nuclear safety and security.

Two-dimensional fluorescence spectroscopy of uranium isotopes in femtosecond laser ablation plumes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Phillips, Mark C.; Brumfield, Brian E.; LaHaye, Nicole L.

We demonstrate measurement of uranium isotopes in femtosecond laser ablation plumes using two-dimensional fluorescence spectroscopy (2DFS). The high-resolution, tunable CW-laser spectroscopy technique clearly distinguishes atomic absorption from 235U and 238U in natural and highly enriched uranium metal samples. We present analysis of spectral resolution and analytical performance of 2DFS as a function of ambient pressure. Simultaneous measurement using time-resolved absorption spectroscopy provides information on temporal dynamics of the laser ablation plume and saturation behavior of fluorescence signals. The rapid, non-contact measurement is promising for in-field, standoff measurements of uranium enrichment for nuclear safety and security applications.

Two-dimensional fluorescence spectroscopy of uranium isotopes in femtosecond laser ablation plumes

DOE PAGES

Phillips, Mark C.; Brumfield, Brian E.; LaHaye, Nicole; ...

2017-06-19

Here, we demonstrate measurement of uranium isotopes in femtosecond laser ablation plumes using two-dimensional fluorescence spectroscopy (2DFS). The high-resolution, tunable CW-laser spectroscopy technique clearly distinguishes atomic absorption from 235U and 238U in natural and highly enriched uranium metal samples. We present analysis of spectral resolution and analytical performance of 2DFS as a function of ambient pressure. Simultaneous measurement using time-resolved absorption spectroscopy provides information on temporal dynamics of the laser ablation plume and saturation behavior of fluorescence signals. The rapid, non-contact measurement is promising for in-field, standoff measurements of uranium enrichment for nuclear safety and security.

Supply of enriched uranium for research reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mueller, H.

1997-08-01

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

Oxidation and crystal field effects in uranium

NASA Astrophysics Data System (ADS)

Tobin, J. G.; Yu, S.-W.; Booth, C. H.; Tyliszczak, T.; Shuh, D. K.; van der Laan, G.; Sokaras, D.; Nordlund, D.; Weng, T.-C.; Bagus, P. S.

2015-07-01

An extensive investigation of oxidation in uranium has been pursued. This includes the utilization of soft x-ray absorption spectroscopy, hard x-ray absorption near-edge structure, resonant (hard) x-ray emission spectroscopy, cluster calculations, and a branching ratio analysis founded on atomic theory. The samples utilized were uranium dioxide (U O2) , uranium trioxide (U O3) , and uranium tetrafluoride (U F4) . A discussion of the role of nonspherical perturbations, i.e., crystal or ligand field effects, will be presented.

On Line Enrichment Monitor (OLEM) UF 6 Tests for 1.5" Sch40 SS Pipe, Revision 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

March-Leuba, José A.; Garner, Jim; Younkin, Jim

As global uranium enrichment capacity under international safeguards expands, the International Atomic Energy Agency (IAEA) is challenged to develop effective safeguards approaches at gaseous centrifuge enrichment plants while working within budgetary constraints. The “Model Safeguards Approach for Gas Centrifuge Enrichment Plants” (GCEPs) developed by the IAEA Division of Concepts and Planning in June 2006, defines the three primary Safeguards objectives to be the timely detection of: 1) diversion of significant quantities of natural (NU), depleted (DU) or low-enriched uranium (LEU) from declared plant flow, 2) facility misuse to produce undeclared LEU product from undeclared feed, and 3) facility misuse tomore » produce enrichments higher than the declared maximum, in particular, highly enriched uranium (HEU). The ability to continuously and independently (i.e. with a minimum of information from the facility operator) monitor not only the uranium mass balance but also the 235U mass balance in the facility could help support all three verification objectives described above. Two key capabilities required to achieve an independent and accurate material balance are 1) continuous, unattended monitoring of in-process UF 6 and 2) monitoring of cylinders entering and leaving the facility. The continuous monitoring of in-process UF 6 would rely on a combination of load-cell monitoring of the cylinders at the feed and withdrawal stations, online monitoring of gas enrichment, and a high-accuracy net weight measurement of the cylinder contents. The Online Enrichment Monitor (OLEM) is the instrument that would continuously measure the time-dependent relative uranium enrichment, E(t), in weight percent 235U, of the gas filling or being withdrawn from the cylinders. The OLEM design concept combines gamma-ray spectrometry using a collimated NaI(Tl) detector with gas pressure and temperature data to calculate the enrichment of the UF 6 gas within the unit header pipe as a function of time. The OLEM components have been tested on ORNL UF 6 flow loop. Data were collected at five different enrichment levels (0.71%, 2.97%, 4.62%, 6.0%, and 93.7%) at several pressure conditions. The test data were collected in the standard OLEM N.4242 file format for each of the conditions with a 10-minute sampling period and then averaged over the span of constant pressures. Analysis of the collected data has provided enrichment constants that can be used for 1.5” stainless steel schedule 40 pipe measurement sites. The enrichment constant is consistent among all the wide range of enrichment levels and pressures used.« less

10 CFR 71.22 - General license: Fissile material.

Code of Federal Regulations, 2011 CFR

2011-01-01

... to obtain the value of X, then the values for the terms in the equation for uranium-233 and plutonium... if: (i) Uranium-233 is present in the package; (ii) The mass of plutonium exceeds 1 percent of the mass of uranium-235; (iii) The uranium is of unknown uranium-235 enrichment or greater than 24 weight...

10 CFR 71.22 - General license: Fissile material.

Code of Federal Regulations, 2012 CFR

2012-01-01

... to obtain the value of X, then the values for the terms in the equation for uranium-233 and plutonium... if: (i) Uranium-233 is present in the package; (ii) The mass of plutonium exceeds 1 percent of the mass of uranium-235; (iii) The uranium is of unknown uranium-235 enrichment or greater than 24 weight...

10 CFR 71.22 - General license: Fissile material.

Code of Federal Regulations, 2014 CFR

2014-01-01

... to obtain the value of X, then the values for the terms in the equation for uranium-233 and plutonium... if: (i) Uranium-233 is present in the package; (ii) The mass of plutonium exceeds 1 percent of the mass of uranium-235; (iii) The uranium is of unknown uranium-235 enrichment or greater than 24 weight...

10 CFR 71.22 - General license: Fissile material.

Code of Federal Regulations, 2010 CFR

2010-01-01

... to obtain the value of X, then the values for the terms in the equation for uranium-233 and plutonium... if: (i) Uranium-233 is present in the package; (ii) The mass of plutonium exceeds 1 percent of the mass of uranium-235; (iii) The uranium is of unknown uranium-235 enrichment or greater than 24 weight...

10 CFR 71.22 - General license: Fissile material.

Code of Federal Regulations, 2013 CFR

2013-01-01

... to obtain the value of X, then the values for the terms in the equation for uranium-233 and plutonium... if: (i) Uranium-233 is present in the package; (ii) The mass of plutonium exceeds 1 percent of the mass of uranium-235; (iii) The uranium is of unknown uranium-235 enrichment or greater than 24 weight...

10 CFR 51.60 - Environmental report-materials licenses.

Code of Federal Regulations, 2011 CFR

2011-01-01

... oil and gas recovery. (vii) Construction and operation of a uranium enrichment facility. (2) Issuance... conversion of uranium hexafluoride pursuant to part 70 of this chapter. (ii) Possession and use of source material for uranium milling or production of uranium hexafluoride pursuant to part 40 of this chapter...

10 CFR 51.60 - Environmental report-materials licenses.

Code of Federal Regulations, 2010 CFR

2010-01-01

... oil and gas recovery. (vii) Construction and operation of a uranium enrichment facility. (2) Issuance... conversion of uranium hexafluoride pursuant to part 70 of this chapter. (ii) Possession and use of source material for uranium milling or production of uranium hexafluoride pursuant to part 40 of this chapter...

10 CFR 51.60 - Environmental report-materials licenses.

Code of Federal Regulations, 2013 CFR

2013-01-01

... oil and gas recovery. (vii) Construction and operation of a uranium enrichment facility. (2) Issuance... conversion of uranium hexafluoride pursuant to part 70 of this chapter. (ii) Possession and use of source material for uranium milling or production of uranium hexafluoride pursuant to part 40 of this chapter...

10 CFR 51.60 - Environmental report-materials licenses.

Code of Federal Regulations, 2014 CFR

2014-01-01

... oil and gas recovery. (vii) Construction and operation of a uranium enrichment facility. (2) Issuance... conversion of uranium hexafluoride pursuant to part 70 of this chapter. (ii) Possession and use of source material for uranium milling or production of uranium hexafluoride pursuant to part 40 of this chapter...

10 CFR 51.60 - Environmental report-materials licenses.

Code of Federal Regulations, 2012 CFR

2012-01-01

... oil and gas recovery. (vii) Construction and operation of a uranium enrichment facility. (2) Issuance... conversion of uranium hexafluoride pursuant to part 70 of this chapter. (ii) Possession and use of source material for uranium milling or production of uranium hexafluoride pursuant to part 40 of this chapter...

49 CFR 173.434 - Activity-mass relationships for uranium and natural thorium.

Code of Federal Regulations, 2011 CFR

2011-10-01

... natural thorium. 173.434 Section 173.434 Transportation Other Regulations Relating to Transportation....434 Activity-mass relationships for uranium and natural thorium. The table of activity-mass relationships for uranium and natural thorium are as follows: Thorium and uranium enrichment 1(Wt% 235 U present...

49 CFR 173.434 - Activity-mass relationships for uranium and natural thorium.

Code of Federal Regulations, 2010 CFR

2010-10-01

... natural thorium. 173.434 Section 173.434 Transportation Other Regulations Relating to Transportation....434 Activity-mass relationships for uranium and natural thorium. The table of activity-mass relationships for uranium and natural thorium are as follows: Thorium and uranium enrichment 1(Wt% 235 U present...

Quantification of 235U and 238U activity concentrations for undeclared nuclear materials by a digital gamma-gamma coincidence spectroscopy.

PubMed

Zhang, Weihua; Yi, Jing; Mekarski, Pawel; Ungar, Kurt; Hauck, Barry; Kramer, Gary H

2011-06-01

The purpose of this study is to investigate the possibility of verifying depleted uranium (DU), natural uranium (NU), low enriched uranium (LEU) and high enriched uranium (HEU) by a developed digital gamma-gamma coincidence spectroscopy. The spectroscopy consists of two NaI(Tl) scintillators and XIA LLC Digital Gamma Finder (DGF)/Pixie-4 software and card package. The results demonstrate that the spectroscopy provides an effective method of (235)U and (238)U quantification based on the count rate of their gamma-gamma coincidence counting signatures. The main advantages of this approach over the conventional gamma spectrometry include the facts of low background continuum near coincident signatures of (235)U and (238)U, less interference from other radionuclides by the gamma-gamma coincidence counting, and region-of-interest (ROI) imagine analysis for uranium enrichment determination. Compared to conventional gamma spectrometry, the method offers additional advantage of requiring minimal calibrations for (235)U and (238)U quantification at different sample geometries. Crown Copyright © 2011. Published by Elsevier Ltd. All rights reserved.

An aerosol particle containing enriched uranium encountered in the remote upper troposphere.

PubMed

Murphy, D M; Froyd, K D; Apel, E; Blake, D; Blake, N; Evangeliou, N; Hornbrook, R S; Peischl, J; Ray, E; Ryerson, T B; Thompson, C; Stohl, A

2018-04-01

We describe a submicron aerosol particle sampled at an altitude of 7 km near the Aleutian Islands that contained a small percentage of enriched uranium oxide. 235 U was 3.1 ± 0.5% of 238 U. During twenty years of aircraft sampling of millions of particles in the global atmosphere, we have rarely encountered a particle with a similarly high content of 238 U and never a particle with enriched 235 U. The bulk of the particle consisted of material consistent with combustion of heavy fuel oil. Analysis of wind trajectories and particle dispersion model results show that the particle could have originated from a variety of areas across Asia. The source of such a particle is unclear, and the particle is described here in case it indicates a novel source where enriched uranium was dispersed. Published by Elsevier Ltd.

Laser removal of loose uranium compound contamination from metal surfaces

NASA Astrophysics Data System (ADS)

Roberts, D. E.; Modise, T. S.

2007-04-01

Pulsed laser removal of surface contamination of uranyl nitrate and uranium dioxide from stainless steel has been studied. Most of the loosely bound contamination has been removed at fluence levels below 0.5 J cm -2, leaving about 5% fixed contamination for uranyl nitrate and 15% for uranium dioxide. Both alpha and beta activities are then sufficiently low that contaminated objects can be taken out of a restricted radiation area for re-use. The ratio of beta to alpha activity is found to be a function of particle size and changes during laser removal. In a separate experiment using technetium-99m, the collection of removed radioactivity in the filter was studied and an inventory made of removed and collected contamination.

Heat deposition analysis for the High Flux Isotope Reactor’s HEU and LEU core models

DOE PAGES

Davidson, Eva E.; Betzler, Benjamin R.; Chandler, David; ...

2017-08-01

The High Flux Isotope Reactor at Oak Ridge National Laboratory is an 85 MW th pressurized light-water-cooled and -moderated flux-trap type research reactor. The reactor is used to conduct numerous experiments, advancing various scientific and engineering disciplines. As part of an ongoing program sponsored by the US Department of Energy National Nuclear Security Administration Office of Material Management and Minimization, studies are being performed to assess the feasibility of converting the reactor’s highly enriched uranium fuel to low-enriched uranium fuel. To support this conversion project, reference models with representative experiment target loading and explicit fuel plate representation were developed andmore » benchmarked for both fuels to (1) allow for consistent comparison between designs for both fuel types and (2) assess the potential impact of low-enriched uranium conversion. These high-fidelity models were used to conduct heat deposition analyses at the beginning and end of the reactor cycle and are presented herein. This article (1) discusses the High Flux Isotope Reactor models developed to facilitate detailed heat deposition analyses of the reactor’s highly enriched and low-enriched uranium cores, (2) examines the computational approach for performing heat deposition analysis, which includes a discussion on the methodology for calculating the amount of energy released per fission, heating rates, power and volumetric heating rates, and (3) provides results calculated throughout various regions of the highly enriched and low-enriched uranium core at the beginning and end of the reactor cycle. These are the first detailed high-fidelity heat deposition analyses for the High Flux Isotope Reactor’s highly enriched and low-enriched core models with explicit fuel plate representation. Lastly, these analyses are used to compare heat distributions obtained for both fuel designs at the beginning and end of the reactor cycle, and they are essential for enabling comprehensive thermal hydraulics and safety analyses that require detailed estimates of the heat source within all of the reactor’s fuel element regions.« less

10 CFR 150.14 - Commission regulatory authority for physical protection.

Code of Federal Regulations, 2012 CFR

2012-01-01

... significance in quantities greater than 15 grams of plutonium or uranium-233 or uranium-235 (enriched to 20 percent or more in the U-235 isotope) or any combination greater than 15 grams when computed by the equation grams=grams uranium-235+grams plutonium+grams uranium-233 shall meet the physical protection...

10 CFR 150.14 - Commission regulatory authority for physical protection.

Code of Federal Regulations, 2010 CFR

2010-01-01

... significance in quantities greater than 15 grams of plutonium or uranium-233 or uranium-235 (enriched to 20 percent or more in the U-235 isotope) or any combination greater than 15 grams when computed by the equation grams=grams uranium-235+grams plutonium+grams uranium-233 shall meet the physical protection...

10 CFR 150.14 - Commission regulatory authority for physical protection.

Code of Federal Regulations, 2011 CFR

2011-01-01

... significance in quantities greater than 15 grams of plutonium or uranium-233 or uranium-235 (enriched to 20 percent or more in the U-235 isotope) or any combination greater than 15 grams when computed by the equation grams=grams uranium-235+grams plutonium+grams uranium-233 shall meet the physical protection...

10 CFR 150.14 - Commission regulatory authority for physical protection.

Code of Federal Regulations, 2013 CFR

2013-01-01

... significance in quantities greater than 15 grams of plutonium or uranium-233 or uranium-235 (enriched to 20 percent or more in the U-235 isotope) or any combination greater than 15 grams when computed by the equation grams=grams uranium-235+grams plutonium+grams uranium-233 shall meet the physical protection...

10 CFR 150.14 - Commission regulatory authority for physical protection.

Code of Federal Regulations, 2014 CFR

2014-01-01

... significance in quantities greater than 15 grams of plutonium or uranium-233 or uranium-235 (enriched to 20 percent or more in the U-235 isotope) or any combination greater than 15 grams when computed by the equation grams=grams uranium-235+grams plutonium+grams uranium-233 shall meet the physical protection...

Equation of state for shock compression of distended solids

NASA Astrophysics Data System (ADS)

Grady, Dennis; Fenton, Gregg; Vogler, Tracy

2014-05-01

Shock Hugoniot data for full-density and porous compounds of boron carbide, silicon dioxide, tantalum pentoxide, uranium dioxide and playa alluvium are investigated for the purpose of equation-of-state representation of intense shock compression. Complications of multivalued Hugoniot behavior characteristic of highly distended solids are addressed through the application of enthalpy-based equations of state of the form originally proposed by Rice and Walsh in the late 1950's. Additive measures of cold and thermal pressure intrinsic to the Mie-Gruneisen EOS framework is replaced by isobaric additive functions of the cold and thermal specific volume components in the enthalpy-based formulation. Additionally, experimental evidence reveals enhancement of shock-induced phase transformation on the Hugoniot with increasing levels of initial distension for silicon dioxide, uranium dioxide and possibly boron carbide. Methods for addressing this experimentally observed feature of the shock compression are incorporated into the EOS model.

Equation of State for Shock Compression of High Distension Solids

NASA Astrophysics Data System (ADS)

Grady, Dennis

2013-06-01

Shock Hugoniot data for full-density and porous compounds of boron carbide, silicon dioxide, tantalum pentoxide, uranium dioxide and playa alluvium are investigated for the purpose of equation-of-state representation of intense shock compression. Complications of multivalued Hugoniot behavior characteristic of highly distended solids are addressed through the application of enthalpy-based equations of state of the form originally proposed by Rice and Walsh in the late 1950's. Additivity of cold and thermal pressure intrinsic to the Mie-Gruneisen EOS framework is replaced by isobaric additive functions of the cold and thermal specific volume components in the enthalpy-based formulation. Additionally, experimental evidence supports acceleration of shock-induced phase transformation on the Hugoniot with increasing levels of initial distention for silicon dioxide, uranium dioxide and possibly boron carbide. Methods for addressing this experimentally observed facet of the shock compression are introduced into the EOS model.

Tags to Track Illicit Uranium and Plutonium

DOE Office of Scientific and Technical Information (OSTI.GOV)

Haire, M. Jonathan; Forsberg, Charles W.

2007-07-01

With the expansion of nuclear power, it is essential to avoid nuclear materials from falling into the hands of rogue nations, terrorists, and other opportunists. This paper examines the idea of detection and attribution tags for nuclear materials. For a detection tag, it is proposed to add small amounts [about one part per billion (ppb)] of {sup 232}U to enriched uranium to brighten its radioactive signature. Enriched uranium would then be as detectable as plutonium and thus increase the likelihood of intercepting illicit enriched uranium. The use of rare earth oxide elements is proposed as a new type of 'attribution'more » tag for uranium and thorium from mills, uranium and plutonium fuels, and other nuclear materials. Rare earth oxides are chosen because they are chemically compatible with the fuel cycle, can survive high-temperature processing operations in fuel fabrication, and can be chosen to have minimal neutronic impact within the nuclear reactor core. The mixture of rare earths and/or rare earth isotopes provides a unique 'bar code' for each tag. If illicit nuclear materials are recovered, the attribution tag can identify the source and lot of nuclear material, and thus help police reduce the possible number of suspects in the diversion of nuclear materials based on who had access. (authors)« less

METHOD OF RECOVERING URANIUM COMPOUNDS

DOEpatents

Poirier, R.H.

1957-10-29

S>The recovery of uranium compounds which have been adsorbed on anion exchange resins is discussed. The uranium and thorium-containing residues from monazite processed by alkali hydroxide are separated from solution, and leached with an alkali metal carbonate solution, whereby the uranium and thorium hydrorides are dissolved. The carbonate solution is then passed over an anion exchange resin causing the uranium to be adsorbed while the thorium remains in solution. The uranium may be recovered by contacting the uranium-holding resin with an aqueous ammonium carbonate solution whereby the uranium values are eluted from the resin and then heating the eluate whereby carbon dioxide and ammonia are given off, the pH value of the solution is lowered, and the uranium is precipitated.

PROCESS OF MAKING A NEUTRONIC REACTOR FUEL ELEMENT COMPOSITION

DOEpatents

Alter, H.W.; Davidson, J.K.; Miller, R.S.; Mewherter, J.L.

1959-01-13

A process is presented for making a ceramic-like material suitable for use as a nuclear fuel. The material consists of a solid solution of plutonium dioxide in uranium dioxide and is produced from a uranyl nitrate -plutonium nitrate solution containing uraniunm and plutonium in the desired ratio. The uranium and plutonium are first precipitated from the solution by addition of NH/ sub 4/OH and the dried precipitate is then calcined at 600 C in a hydrogen atmosphere to yield the desired solid solution of PuO/sub 2/ in UO/sub 2/.

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

Code of Federal Regulations, 2011 CFR

2011-01-01

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

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

Code of Federal Regulations, 2010 CFR

2010-01-01

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

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

Federal Register 2010, 2011, 2012, 2013, 2014

2013-11-07

... in U.S. customs territory, and (ii) are re-exported within eighteen (18) months of entry of the low... extend the deadline for re-exportation of this sole entry of low-enriched uranium. The Department determines that the deadline for re-exportation of this sole entry is November 1, 2015, and that this will be...

Routine inspection effort required for verification of a nuclear material production cutoff convention

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dougherty, D.; Fainberg, A.; Sanborn, J.

On 27 September 1993, President Clinton proposed {open_quotes}... a multilateral convention prohibiting the production of highly enriched uranium or plutonium for nuclear explosives purposes or outside of international safeguards.{close_quotes} The UN General Assembly subsequently adopted a resolution recommending negotiation of a non-discriminatory, multilateral, and internationally and effectively verifiable treaty (hereinafter referred to as {open_quotes}the Cutoff Convention{close_quotes}) banning the production of fissile material for nuclear weapons. The matter is now on the agenda of the Conference on Disarmament, although not yet under negotiation. This accord would, in effect, place all fissile material (defined as highly enriched uranium and plutonium) produced aftermore » entry into force (EIF) of the accord under international safeguards. {open_quotes}Production{close_quotes} would mean separation of the material in question from radioactive fission products, as in spent fuel reprocessing, or enrichment of uranium above the 20% level, which defines highly enriched uranium (HEU). Facilities where such production could occur would be safeguarded to verify that either such production is not occurring or that all material produced at these facilities is maintained under safeguards.« less

Uranium enrichment in lacustrine oil source rocks of the Chang 7 member of the Yanchang Formation, Erdos Basin, China

NASA Astrophysics Data System (ADS)

Yang, Hua; Zhang, Wenzheng; Wu, Kai; Li, Shanpeng; Peng, Ping'an; Qin, Yan

2010-09-01

The oil source rocks of the Chang 7 member of the Yanchang Formation in the Erdos Basin were deposited during maximum lake extension during the Late Triassic and show a remarkable positive uranium anomaly, with an average uranium content as high as 51.1 μg/g. Uranium is enriched together with organic matter and elements such as Fe, S, Cu, V and Mo in the rocks. The detailed biological markers determined in the Chang 7 member indicate that the lake water column was oxidizing during deposition of the Chang 7 member. However, redox indicators for sediments such as S 2- content, V/Sc and V/(V + Ni) ratios demonstrate that it was a typical anoxic diagenetic setting. The contrasted redox conditions between the water column and the sediment with a very high content of organic matter provided favorable physical and chemical conditions for syngenetic uranium enrichment in the oil source rocks of the Chang 7 member. Possible uranium sources may be the extensive U-rich volcanic ash that resulted from contemporaneous volcanic eruption and uranium material transported by hydrothermal conduits into the basin. The uranium from terrestrial clastics was unlike because uranium concentration was not higher in the margin area of basin where the terrestrial material input was high. As indicated by correlative analysis, the oil source rocks of the Chang 7 member show high gamma-ray values for radioactive well log data that reflect a positive uranium anomaly and are characterized by high resistance, low electric potential and low density. As a result, well log data can be used to identify positive uranium anomalies and spatial distribution of the oil source rocks in the Erdos Basin. The estimation of the total uranium reserves in the Chang 7 member attain 0.8 × 10 8 t.

76 FR 34103 - In the Matter of Areva Enrichment Services, LLC (Eagle Rock Enrichment Facility); Notice of...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-06-10

.... 10-899-02-ML-BD01] In the Matter of Areva Enrichment Services, LLC (Eagle Rock Enrichment Facility... gas centrifuge uranium enrichment facility--denoted as the Eagle Rock Enrichment Facility (EREF)--in... Information for Contention Preparation; In the Matter of Areva Enrichment Services, LLC (Eagle Rock Enrichment...

METHOD OF PRODUCING URANIUM METAL BY ELECTROLYSIS

DOEpatents

Piper, R.D.

1962-09-01

A process is given for making uranium metal from oxidic material by electrolytic deposition on the cathode. The oxidic material admixed with two moles of carbon per one mole of uranium dioxide forms the anode, and the electrolyte is a mixture of from 40 to 75% of calcium fluoride or barium fluoride, 15 to 45% of uranium tetrafluoride, and from 10 to 20% of lithium fluoride or magnesium fluoride; the temperature of the electrolyte is between 1150 and 1175 deg C. (AEC)

Loading blended, low-enriched uranium fuel in browns ferry units 2 and 3

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, C.; Eichenberg, T.; Haun, J.

2006-07-01

This paper summarizes fuel and cycle design results for the Tennessee Valley Authority (TVA) / Dept. of Energy (DOE) program to burn blended, low-enriched uranium (BLEU) material in the Browns Ferry Nuclear Units 2 and 3. The BLEU material typically has about 60 times the allowed limit of U-236 in what would be defined as commercial, i.e., virgin, uranium. U-236 in particular is a strong neutron absorber. Also included is a comparison of cycles using commercial uranium versus BLEU to determine the impact on key core design parameters of the high U-236 content in the BLEU. Finally, there is amore » short discussion of the economic advantages of BLEU fuel. (authors)« less

Experimental investigation of neutronic characteristics of the IR-8 reactor to confirm the results of calculations by MCU-PTR code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Surkov, A. V., E-mail: surkov.andrew@gmail.com; Kochkin, V. N.; Pesnya, Yu. E.

2015-12-15

A comparison of measured and calculated neutronic characteristics (fast neutron flux and fission rate of {sup 235}U) in the core and reflector of the IR-8 reactor is presented. The irradiation devices equipped with neutron activation detectors were prepared. The determination of fast neutron flux was performed using the {sup 54}Fe (n, p) and {sup 58}Ni (n, p) reactions. The {sup 235}U fission rate was measured using uranium dioxide with 10% enrichment in {sup 235}U. The determination of specific activities of detectors was carried out by measuring the intensity of characteristic gamma peaks using the ORTEC gamma spectrometer. Neutron fields inmore » the core and reflector of the IR-8 reactor were calculated using the MCU-PTR code.« less

From the Lab to the real world : sources of error in UF {sub 6} gas enrichment monitoring

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lombardi, Marcie L.

2012-03-01

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 to an improved safeguards model that calls for new technologies to help keep up with the increasing size and complexity of today’s gas centrifuge enrichment plants (GCEPs). One of the primary goals of the IAEA is to detect the production of uranium at levels greater than those an enrichment facility may have declared. In order to accomplish this goal, new enrichment monitors need to be as accurate as possible. This dissertation will look at themore » Advanced Enrichment Monitor (AEM), a new enrichment monitor designed at Los Alamos National Laboratory. Specifically explored are various factors that could potentially contribute to errors in a final enrichment determination delivered by the AEM. There are many factors that can cause errors in the determination of uranium hexafluoride (UF{sub 6}) gas enrichment, especially during the period when the enrichment is being measured in an operating GCEP. To measure enrichment using the AEM, a passive 186-keV (kiloelectronvolt) measurement is used to determine the {sup 235}U content in the gas, and a transmission measurement or a gas pressure reading is used to determine the total uranium content. A transmission spectrum is generated using an x-ray tube and a “notch” filter. In this dissertation, changes that could occur in the detection efficiency and the transmission errors that could result from variations in pipe-wall thickness will be explored. Additional factors that could contribute to errors in enrichment measurement will also be examined, including changes in the gas pressure, ambient and UF{sub 6} temperature, instrumental errors, and the effects of uranium deposits on the inside of the pipe walls will be considered. The sensitivity of the enrichment calculation to these various parameters will then be evaluated. Previously, UF{sub 6} gas enrichment monitors have required empty pipe measurements to accurately determine the pipe attenuation (the pipe attenuation is typically much larger than the attenuation in the gas). This dissertation reports on a method for determining the thickness of a pipe in a GCEP when obtaining an empty pipe measurement may not be feasible. This dissertation studies each of the components that may add to the final error in the enrichment measurement, and the factors that were taken into account to mitigate these issues are also detailed and tested. The use of an x-ray generator as a transmission source and the attending stability issues are addressed. Both analytical calculations and experimental measurements have been used. For completeness, some real-world analysis results from the URENCO Capenhurst enrichment plant have been included, where the final enrichment error has remained well below 1% for approximately two months.« less

Uranium and Thorium

ERIC Educational Resources Information Center

Finch, Warren I.

1978-01-01

The results of President Carter's policy on non-proliferation of nuclear weapons are expected to slow the growth rate in energy consumption, put the development of the breeder reactor in question, halt plans to reprocess and recycle uranium and plutonium, and expand facilities to supply enriched uranium. (Author/MA)

Design of a Uranium Dioxide Spheroidization System

NASA Technical Reports Server (NTRS)

Cavender, Daniel P.; Mireles, Omar R.; Frendi, Abdelkader

2013-01-01

The plasma spheroidization system (PSS) is the first process in the development of tungsten-uranium dioxide (W-UO2) fuel cermets. The PSS process improves particle spherocity and surface morphology for coating by chemical vapor deposition (CVD) process. Angular fully dense particles melt in an argon-hydrogen plasma jet at between 32-36 kW, and become spherical due to surface tension. Surrogate CeO2 powder was used in place of UO2 for system and process parameter development. Particles range in size from 100 - 50 microns in diameter. Student s t-test and hypothesis testing of two proportions statistical methods were applied to characterize and compare the spherocity of pre and post process powders. Particle spherocity was determined by irregularity parameter. Processed powders show great than 800% increase in the number of spherical particles over the stock powder with the mean spherocity only mildly improved. It is recommended that powders be processed two-three times in order to reach the desired spherocity, and that process parameters be optimized for a more narrow particles size range. Keywords: spherocity, spheroidization, plasma, uranium-dioxide, cermet, nuclear, propulsion

Returning HEU Fuel from the Czech Republic to Russia

DOE Office of Scientific and Technical Information (OSTI.GOV)

Michael Tyacke; Dr. Igor Bolshinsky

In December 1999, representatives from the United States, Russian Federation, and International Atomic Energy Agency began working on a program to return Russian supplied, highly enriched, uranium fuel stored at foreign research reactors to Russia. Now, under the Global Threat Reduction Initiative’s Russian Research Reactor Fuel Return Program, this effort has repatriated over 800 kg of highly enriched uranium to Russia from over 10 countries. In May 2004, the “Agreement Between the Government of the United States of America and the Government of the Russian Federation Concerning Cooperation for the Transfer of Russian Produced Research Reactor Nuclear Fuel to themore » Russian Federation” was signed. This agreement provides legal authority for the Russian Research Reactor Fuel Return Program and establishes parameters whereby eligible countries may return highly enriched uranium spent and fresh fuel assemblies and other fissile materials to Russia. On December 8, 2007, one of the largest shipments of highly 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. In February 2003, Russian Research Reactor Fuel Return Program representatives met with the Nuclear Research Institute in Rež, Czech Republic, and discussed the return of their highly enriched uranium spent nuclear fuel to the Russian Federation for reprocessing. Nearly 5 years later, the shipment was made. This article discusses the planning, preparations, coordination, and cooperation required to make this important international shipment.« less

SULPHUR DIOXIDE LEACHING OF URANIUM CONTAINING MATERIAL

DOEpatents

Thunaes, A.; Rabbits, F.T.; Hester, K.D.; Smith, H.W.

1958-12-01

A process is described for extracting uranlum from uranium containing material, such as a low grade pitchblende ore, or mill taillngs, where at least part of the uraniunn is in the +4 oxidation state. After comminuting and magnetically removing any entrained lron particles the general material is made up as an aqueous slurry containing added ferric and manganese salts and treated with sulfur dioxide and aeration to an extent sufficient to form a proportion of oxysulfur acids to give a pH of about 1 to 2 but insufficient to cause excessive removal of the sulfur dioxide gas. After separating from the solids, the leach solution is adjusted to a pH of about 1.25, then treated with metallic iron in the presence of a precipitant such as a soluble phosphate, arsonate, or fluoride.

Alternative Anodes for the Electrolytic Reduction of Uranium Dioxide

NASA Astrophysics Data System (ADS)

Merwin, Augustus

Reprocessing of spent nuclear fuel is an essential step in closing the nuclear fuel cycle. In order to consume current stockpiles, ceramic uranium dioxide spent nuclear fuel will be subjected to an electrolytic reduction process. The current reduction process employs a platinum anode and a stainless steel alloy 316 cathode in a molten salt bath consisting of LiCl-2wt% Li 2O and occurs at 700°C. A major shortcoming of the existing process is the degradation of the platinum anode under the severely oxidizing conditions encountered during electrolytic reduction. This work investigates alternative anode materials for the electrolytic reduction of uranium oxide. The high temperature and extreme oxidizing conditions encountered in these studies necessitated a unique set of design constraints on the system. Thus, a customized experimental apparatus was designed and constructed. The electrochemical experiments were performed in an electrochemical reactor placed inside a furnace. This entire setup was housed inside a glove box, in order to maintain an inert atmosphere. This study investigates alternative anode materials through accelerated corrosion testing. Surface morphology was studied using scanning electron microscopy. Surface chemistry was characterized using energy dispersive spectroscopy and Raman spectroscopy. Electrochemical behavior of candidate materials was evaluated using potentiodynamic polarization characteristics. After narrowing the number of candidate electrode materials, ferrous stainless steel alloy 316, nickel based Inconel 718 and elemental tungsten were chosen for further investigation. Of these materials only tungsten was found to be sufficiently stable at the anodic potential required for electrolysis of uranium dioxide in molten salt. The tungsten anode and stainless steel alloy 316 cathode electrode system was studied at the required reduction potential for UO2 with varying lithium oxide concentrations. Electrochemical impedance spectroscopy showed mixed (kinetic and diffusion) control and an overall low impedance due to extreme corrosion. It was observed that tungsten is sufficiently stable in LiCl - 2wt% Li 2O at 700°C at the required anodic potential for the reduction of uranium oxide. This study identifies tungsten to be a superior anode material to platinum for the electrolytic reduction of uranium oxide, both in terms of superior corrosion behavior and reduced cost, and thus recommends that tungsten be further investigated as an alternative anode for the electrolytic reduction of uranium dioxide.

Development of Novel Porous Sorbents for Extraction of Uranium from Seawater

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lin, Wenbin

Climate disruption is one of the greatest crises the global community faces in the 21st century. Alarming increases in CO 2, NO, SO 2 and particulate matter levels will have catastrophic consequences on the environment, food supplies, and human health if no action is taken to lessen their worldwide prevalence. Nuclear energy remains the only mature technology capable of continuous base-load power generation with ultralow carbon dioxide, nitric oxide, and sulfur dioxide emissions. Over the lifetime of the technology, nuclear energy outputs less than 1.5% the carbon dioxide emissions per gigawatt hour relative to coal—about as much as onshore windmore » power.1 However, in order for nuclear energy to be considered a viable option in the future, a stable supply of uranium must be secured. Current estimates suggest there is less than 100 years’ worth of uranium left in terrestrial ores (6.3 million tons) if current consumption levels remain unchanged.2 It is likely, however, that demand for nuclear fuel will rise as a direct consequence of the ratification of global climate accords. The oceans, containing approximately 4.5 billion tons of uranium (U) at a uniform concentration of ~3 ppb, represent a virtually limitless supply of this resource.3 Development of technologies to recover uranium from seawater would greatly improve the U resource availability, providing a U price ceiling for the current generation and sustaining the nuclear fuel supply for future generations. Several methods have been previously evaluated for uranium sequestration including solvent extraction, ion exchange, flotation, biomass collection, and adsorption; however, none have been found to be suitable for reasons including cost effectiveness, long term stability, and selectivity.4,5 While polymer beads and fibers have been functionalized with amidoxime functional groups to afford U adsorption capacities as high as 1.5 g U/kg,6 further discoveries are needed to make uranium extraction from seawater economically feasible.« less

Sandstone type uranium deposits in the Ordos Basin, Northwest China: A case study and an overview

NASA Astrophysics Data System (ADS)

Akhtar, Shamim; Yang, Xiaoyong; Pirajno, Franco

2017-09-01

This paper provides a comprehensive review on studies of sandstone type uranium deposits in the Ordos Basin, Northwest China. As the second largest sedimentary basin, the Ordos Basin has great potential for targeting sandstone type U mineralization. The newly found and explored Dongsheng and Diantou sandstone type uranium deposits are hosted in the Middle Jurassic Zhilou Formation. A large number of investigations have been conducted to trace the source rock compositions and relationship between lithic subarkose sandstone host rock and uranium mineralization. An optical microscopy study reveals two types of alteration associated with the U mineralization: chloritization and sericitization. Some unusual mineral structures, with compositional similarity to coffinite, have been identified in a secondary pyrite by SEM These mineral phases are proposed to be of bacterial origin, following high resolution mapping of uranium minerals and trace element determinations in situ. Moreover, geochemical studies of REE and trace elements constrained the mechanism of uranium enrichment, displaying LREE enrichment relative to HREE. Trace elements such as Pb, Mo and Ba have a direct relationship with uranium enrichment and can be used as index for mineralization. The source of uranium ore forming fluids and related geological processes have been studied using H, O and C isotope systematics of fluid inclusions in quartz veins and the calcite cement of sandstone rocks hosting U mineralization. Both H and O isotopic compositions of fluid inclusions reveal that ore forming fluids are a mixture of meteoric water and magmatic water. The C and S isotopes of the cementing material of sandstone suggest organic origin and bacterial sulfate reduction (BSR), providing an important clue for U mineralization. Discussion of the ore genesis shows that the greenish gray sandstone plays a crucial role during processes leading to uranium mineralization. Consequently, an oxidation-reduction model for sandstone-type uranium deposit is proposed, which can elucidate the source of uranium in the deposits of the Ordos Basin, based on the role of organic materials and sulfate reducing bacteria. We discuss the mechanism of uranium deposition responsible for the genesis of these large sandstone type uranium deposits in this unique sedimentary basin.

Covalency in oxidized uranium

NASA Astrophysics Data System (ADS)

Tobin, J. G.; Yu, S.-W.; Qiao, R.; Yang, W. L.; Booth, C. H.; Shuh, D. K.; Duffin, A. M.; Sokaras, D.; Nordlund, D.; Weng, T.-C.

2015-07-01

Using x-ray emission spectroscopy and absorption spectroscopy, it has been possible to directly access the states in the unoccupied conduction bands that are involved with 5 f and 6 d covalency in oxidized uranium. By varying the oxidizing agent, the degree of 5 f covalency can be manipulated and monitored, clearly and irrevocably establishing the importance of 5 f covalency in the electronic structure of the key nuclear fuel, uranium dioxide.

Covalency in oxidized uranium

DOE PAGES

Tobin, J. G.; Yu, S. -W.; Qiao, R.; ...

2015-07-01

Here, using x-ray emission spectroscopy and absorption spectroscopy, it has been possible to directly access the states in the unoccupied conduction bands that are involved with 5f and 6d covalency in oxidized uranium. By varying the oxidizing agent, the degree of 5f covalency can be manipulated and monitored, clearly and irrevocably establishing the importance of 5f covalency in the electronic structure of the key nuclear fuel, uranium dioxide.

Removing oxygen from a solvent extractant in an uranium recovery process

DOEpatents

Hurst, Fred J.; Brown, Gilbert M.; Posey, Franz A.

1984-01-01

An improvement in effecting uranium recovery from phosphoric acid solutions is provided by sparging dissolved oxygen contained in solutions and solvents used in a reductive stripping stage with an effective volume of a nonoxidizing gas before the introduction of the solutions and solvents into the stage. Effective volumes of nonoxidizing gases, selected from the group consisting of argon, carbon dioxide, carbon monoxide, helium, hydrogen, nitrogen, sulfur dioxide, and mixtures thereof, displace oxygen from the solutions and solvents thereby reduce deleterious effects of oxygen such as excessive consumption of elemental or ferrous and accumulation of complex iron phosphates or cruds.

Aerodynamic levitator for in situ x-ray structure measurements on high temperature and molten nuclear fuel materials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Weber, J. K. R.; Alderman, O. L. G.; Advanced Photon Source, Argonne National Laboratory, Argonne, Illinois 60439

2016-07-15

An aerodynamic levitator with carbon dioxide laser beam heating was integrated with a hermetically sealed controlled atmosphere chamber and sample handling mechanism. The system enabled containment of radioactive samples and control of the process atmosphere chemistry. The chamber was typically operated at a pressure of approximately 0.9 bars to ensure containment of the materials being processed. Samples 2.5-3 mm in diameter were levitated in flowing gas to achieve containerless conditions. Levitated samples were heated to temperatures of up to 3500 °C with a partially focused carbon dioxide laser beam. Sample temperature was measured using an optical pyrometer. The sample environment wasmore » integrated with a high energy (100 keV) x-ray synchrotron beamline to enable in situ structure measurements to be made on levitated samples as they were heated, melted, and supercooled. The system was controlled from outside the x-ray beamline hutch by using a LabVIEW program. Measurements have been made on hot solid and molten uranium dioxide and binary uranium dioxide-zirconium dioxide compositions.« less

Aerodynamic levitator for in situ x-ray structure measurements on high temperature and molten nuclear fuel materials

DOE PAGES

Weber, J. K. R.; Tamalonis, A.; Benmore, C. J.; ...

2016-07-01

We integrated an aerodynamic levitator with carbon dioxide laser beam heating with a hermetically sealed controlled atmosphere chamber and sample handling mechanism. The system enabled containment of radioactive samples and control of the process atmosphere chemistry. Furthermore, the chamber was typically operated at a pressure of approximately 0.9 bars to ensure containment of the materials being processed. Samples 2.5-3 mm in diameter were levitated in flowing gas to achieve containerless conditions. Levitated samples were heated to temperatures of up to 3500 °C with a partially focused carbon dioxide laser beam. Sample temperature was measured using an optical pyrometer. The samplemore » environment was integrated with a high energy (100 keV) x-ray synchrotron beamline to enable in situ structure measurements to be made on levitated samples as they were heated, melted, and supercooled. Our system was controlled from outside the x-ray beamline hutch by using a LabVIEW program. Measurements have been made on hot solid and molten uranium dioxide and binary uranium dioxide-zirconium dioxide compositions.« less

24. VIEW OF THE SECOND FLOOR PLAN. ENRICHED URANIUM AND ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

24. VIEW OF THE SECOND FLOOR PLAN. ENRICHED URANIUM AND STAINLESS STEEL WEAPONS COMPONENT PRODUCTION-RELATED ACTIVITIES OCCURRED PRIMARILY ON THE SECOND FLOOR. THE ORIGINAL DRAWING HAS BEEN ARCHIVED ON MICROFILM. THE DRAWING WAS REPRODUCED AT THE BEST QUALITY POSSIBLE. LETTERS AND NUMBERS IN THE CIRCLES INDICATE FOOTER AND/OR COLUMN LOCATIONS. - Rocky Flats Plant, General Manufacturing, Support, Records-Central Computing, Southern portion of Plant, Golden, Jefferson County, CO

Future World of Illicit Nuclear Trade: Mitigating the Threat

DTIC Science & Technology

2013-07-29

uranium with lasers that is similar to MLIS. 3 Most of the equipment, including four carbon monoxide lasers and vacuum chambers, was delivered. But...Centrifuge Facility 43 Figure 10: Centrifuge Output vs. Goods Required 44 3b Digging Deeper: Laser Enrichment of Uranium 47 Box 3...Major Foreign Assistance to Iran’s Pre-2004 Laser Enrichment Program 50 4. Key Information: The Special Challenge of the Spread of Classified 53

10 CFR 75.4 - Definitions.

Code of Federal Regulations, 2012 CFR

2012-01-01

... uranium or enriching uranium in the isotope 235, zirconium tubes, heavy water or deuterium, nuclear-grade..., irradiated fuel element chopping machines, and hot cells. Nuclear fuel cycle-related research and development...

HEU Holdup Measurements in 321-M B and Spare U-Al Casting Furnaces

DOE Office of Scientific and Technical Information (OSTI.GOV)

Salaymeh, S.R.

The Analytical Development Section of Savannah River Technology Center (SRTC) was requested by the Facilities Decontamination Division (FDD) to determine the holdup of enriched uranium in the 321-M facility as part of an overall deactivation project of the facility. The 321-M facility was used to fabricate enriched uranium fuel assemblies, lithium-aluminum target tubes, neptunium assemblies, and miscellaneous components for the production reactors. This report covers holdup measurements in two uranium aluminum alloy (U-Al) casting furnaces. Our results indicate an upper limit of 235U content for the B and Spare furnaces of 51 and 67 g respectively. This report discusses themore » methodology, non-destructive assay (NDA) measurements, and results of the uranium holdup on the two furnaces.« less

10 CFR 70.59 - Effluent monitoring reporting requirements.

Code of Federal Regulations, 2011 CFR

2011-01-01

... fabrication, scrap recovery, conversion of uranium hexafluoride, or in a uranium enrichment facility shall... this specifically. On the basis of these reports and any additional information the Commission may...

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

Federal Register 2010, 2011, 2012, 2013, 2014

2011-01-04

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Luther, Erik Paul; Leckie, Rafael M.; Dombrowski, David E.

This supplemental report describes fuel fabrication efforts conducted for the Idaho National Laboratory Trade Study for the TREAT Conversion project that is exploring the replacement of the HEU (Highly Enriched Uranium) fuel core of the TREAT reactor with LEU (Low Enriched Uranium) fuel. Previous reports have documented fabrication of fuel by the “upgrade” process developed at Los Alamos National Laboratory. These experiments supplement an earlier report that describes efforts to increase the graphite content of extruded fuel and minimize cracking.

Pakistan’s Nuclear Weapons: Proliferation and Security Issues

DTIC Science & Technology

2009-12-09

Nuclear Terrorism in Pakistan: Sabotage of a Spent Fuel Cask or a Commercial Irradiation Source in Transport ,” in Pakistan’s Nuclear Future, 2008...gave additional urgency to the program. Pakistan produced fissile material for its nuclear weapons using gas-centrifuge-based uranium enrichment...technology, which it mastered by the mid-1980s. Highly-enriched uranium (HEU) is one of two types of fissile material used in nuclear weapons; the other

Israel: Background and U.S. Relations

DTIC Science & Technology

2013-11-01

material that could be used for nuclear weapons—apparently adding to existing Israeli concerns regarding Iranian uranium enrichment. The reactor under...York, September 24, 2013. 45 Walter Russell Mead, “Threading the Needle,” blogs.the-american-interest.com, October 25, 2013. 46 Israel Prime...civil war,” haaretz.com, September 17, 2013. 59 “‘Israel will not accept deal that allows Iran to enrich uranium ,’” israelhayom.com, October 23, 2013

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Michael Tyacke; Dr. Igor Bolshinsky; Jeff Chamberlin

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

Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core.

PubMed

Lashkari, A; Khalafi, H; Kazeminejad, H

2013-05-01

In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR_PC package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change.

Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core

PubMed Central

Lashkari, A.; Khalafi, H.; Kazeminejad, H.

2013-01-01

In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR_PC package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change. PMID:24976672

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

Federal Register 2010, 2011, 2012, 2013, 2014

2010-10-13

... Evaluation Report; AREVA Enrichment Services LLC, Eagle Rock Enrichment Facility, Bonneville County, ID... report. FOR FURTHER INFORMATION CONTACT: Breeda Reilly, Senior Project Manager, Advanced Fuel Cycle, Enrichment, and Uranium Conversion, Division of Fuel Cycle Safety and Safeguards, Office of Nuclear Material...

RERTR 2009 (Reduced Enrichment for Research and Test Reactors)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Totev, T.; Stevens, J.; Kim, Y. S.

2010-03-01

The U.S. Department of Energy/National Nuclear Security Administration's Office of Global Threat Reduction in cooperation with the China Atomic Energy Authority and International Atomic Energy Agency hosted the 'RERTR 2009 International Meeting on Reduced Enrichment for Research and Test Reactors.' The meeting was organized by Argonne National Laboratory, China Institute of Atomic Energy and Idaho National Laboratory and was held in Beijing, China from November 1-5, 2009. This was the 31st annual meeting in a series on the same general subject regarding the conversion of reactors within the Global Threat Reduction Initiative (GTRI). The Reduced Enrichment for Research and Testmore » Reactors (RERTR) Program develops technology necessary to enable the conversion of civilian facilities using high enriched uranium (HEU) to low enriched uranium (LEU) fuels and targets.« less

Nuclear energy in Europe: uranium flow modeling and fuel cycle scenario trade-offs from a sustainability perspective.

PubMed

Tendall, Danielle M; Binder, Claudia R

2011-03-15

The European nuclear fuel cycle (covering the EU-27, Switzerland and Ukraine) was modeled using material flow analysis (MFA).The analysis was based on publicly available data from nuclear energy agencies and industries, national trade offices, and nongovernmental organizations. Military uranium was not considered due to lack of accessible data. Nuclear fuel cycle scenarios varying spent fuel reprocessing, depleted uranium re-enrichment, enrichment assays, and use of fast neutron reactors, were established. They were then assessed according to environmental, economic and social criteria such as resource depletion, waste production, chemical and radiation emissions, costs, and proliferation risks. The most preferable scenario in the short term is a combination of reduced tails assay and enrichment grade, allowing a 17.9% reduction of uranium demand without significantly increasing environmental, economic, or social risks. In the long term, fast reactors could theoretically achieve a 99.4% decrease in uranium demand and nuclear waste production. However, this involves important costs and proliferation risks. Increasing material efficiency is not systematically correlated with the reduction of other risks. This suggests that an overall optimization of the nuclear fuel cycle is difficult to obtain. Therefore, criteria must be weighted according to stakeholder interests in order to determine the most sustainable solution. This paper models the flows of uranium and associated materials in Europe, and provides a decision support tool for identifying the trade-offs of the alternative nuclear fuel cycles considered.

First-principles study on oxidation effects in uranium oxides and high-pressure high-temperature behavior of point defects in uranium dioxide

NASA Astrophysics Data System (ADS)

Geng, Hua Y.; Song, Hong X.; Jin, K.; Xiang, S. K.; Wu, Q.

2011-11-01

Formation Gibbs free energy of point defects and oxygen clusters in uranium dioxide at high-pressure high-temperature conditions are calculated from first principles, using the LSDA+U approach for the electronic structure and the Debye model for the lattice vibrations. The phonon contribution on Frenkel pairs is found to be notable, whereas it is negligible for the Schottky defect. Hydrostatic compression changes the formation energies drastically, making defect concentrations depend more sensitively on pressure. Calculations show that, if no oxygen clusters are considered, uranium vacancy becomes predominant in overstoichiometric UO2 with the aid of the contribution from lattice vibrations, while compression favors oxygen defects and suppresses uranium vacancy greatly. At ambient pressure, however, the experimental observation of predominant oxygen defects in this regime can be reproduced only in a form of cuboctahedral clusters, underlining the importance of defect clustering in UO2+x. Making use of the point defect model, an equation of state for nonstoichiometric oxides is established, which is then applied to describe the shock Hugoniot of UO2+x. Furthermore, the oxidization and compression behavior of uranium monoxide, triuranium octoxide, uranium trioxide, and a series of defective UO2 at 0 K are investigated. The evolution of mechanical properties and electronic structures with an increase of the oxidation degree are analyzed, revealing the transition of the ground state of uranium oxides from metallic to Mott insulator and then to charge-transfer insulator due to the interplay of strongly correlated effects of 5f orbitals and the shift of electrons from uranium to oxygen atoms.

Boron coating on boron nitride coated nuclear fuels by chemical vapor deposition

NASA Astrophysics Data System (ADS)

Durmazuçar, Hasan H.; Gündüz, Güngör

2000-12-01

Uranium dioxide-only and uranium dioxide-gadolinium oxide (5% and 10%) ceramic nuclear fuel pellets which were already coated with boron nitride were coated with thin boron layer by chemical vapor deposition to increase the burn-up efficiency of the fuel during reactor operation. Coating was accomplished from the reaction of boron trichloride with hydrogen at 1250 K in a tube furnace, and then sintering at 1400 and 1525 K. The deposited boron was identified by infrared spectrum. The morphology of the coating was studied by using scanning electron microscope. The plate, grainy and string (fiber)-like boron structures were observed.

Method for oxygen reduction in a uranium-recovery process. [US DOE patent application

DOEpatents

Hurst, F.J.; Brown, G.M.; Posey, F.A.

1981-11-04

An improvement in effecting uranium recovery from phosphoric acid solutions is provided by sparging dissolved oxygen contained in solutions and solvents used in a reductive stripping stage with an effective volume of a nonoxidizing gas before the introduction of the solutions and solvents into the stage. Effective volumes of nonoxidizing gases, selected from the group consisting of argon, carbon dioxide, carbon monoxide, helium, hydrogen, nitrogen, sulfur dioxide, and mixtures thereof, displace oxygen from the solutions and solvents thereby reduce deleterious effects of oxygen such as excessive consumption of elemental or ferrous iron and accumulation of complex iron phosphates or cruds.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Percher, Catherine G.

2011-08-08

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dale, Gregory E.

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

JPRS Report Science and Technology, Japan: Atomic Energy Society 1989 Annual Meeting.

DTIC Science & Technology

1989-10-13

Control Rod Hole in VHTRC-1 Core [F, Akino, T, Yamane, et al.] ,,, 5 Measurement of MEU [Medium Enriched Uranium ] Fuel Element Characteristics in...K. Yoshida, K. Kobayashi, I. Kimura , C. Yamanaka, and S. Nakai, Laser Laboratory,, Osaka University. Nuclear Reactor Laboratory, Kyoto University...1 core loaded with 278 fuel rods (4 percent enriched uranium ). The PNS target was placed at the back center of the 1/2 assembly on the fixed side

Utilization of non-weapons-grade plutonium and highly enriched uranium with breeding of the 233U isotope in the VVER reactors using thorium and heavy water

NASA Astrophysics Data System (ADS)

Marshalkin, V. E.; Povyshev, V. M.

2015-12-01

A method for joint utilization of non-weapons-grade plutonium and highly enriched uranium in the thorium-uranium—plutonium oxide fuel of a water-moderated reactor with a varying water composition (D2O, H2O) is proposed. The method is characterized by efficient breeding of the 233U isotope and safe reactor operation and is comparatively simple to implement.

Vector representation as a tool for detecting characteristic uranium peaks

NASA Astrophysics Data System (ADS)

Forney, Anne Marie

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 accomplished through the isolation of two regions of interest around the uranium-235 (235U) gamma emission at 186 keV and the uranium-238 (238U) gamma emission at 1001 keV. The uranium 186 keV peak is used as a meter for uranium enrichment, and events from this emission occurred more frequently with the increase of the 235U composition. Spectra were taken with the use of a high purity germanium detector in series with a multi-channel analyzer (MCA) and Maestro 32, a MCA emulator and spectral software. The vector representation method was used to compare two spectra by taking their dot product. The output from this method is an angle, which represents the similarity and contrast between the two spectra. When the angle is close to zero the spectra are similar, and as the angle approaches 90 degrees the spectral agreement decreases. The angles were calculated and compared in Microsoft Excel. A 49 % enriched uranyl acetate source containing both gamma emissions from 235U and 238U was used as a reference source to which all spectra were compared. Two other uranium sources were used within this project: a 100.2 nCi highly-enriched uranium source with 97.7 % 235U by weight, and a piece of uranium ore with an approximate exposure rate of 0.2 mR/h (51.5 nC/kg/h) at 1 cm. These two uranium sources provided different ratios of 235U to 238U, leading to different ratios of the 186 keV and 1001 keV peaks. To test the limits of the vector representation method, various source configurations were used. These included placing the source directly on top of the detector, using two distances for the source from the detector, using the source in addition to cobalt-60, and finally two distances for the source from the detector with a one centimeter lead shield. The two distances from the detector without the shielding were 1.3 inches (3.30 cm) and 1 foot (30.48 cm). In the cases using lead shielding, in the first geometry, the source was placed directly on the lead shielding and in the second geometry, the source was placed a foot above the lead shielding and detector. Vector representation output angles higher than a value of 40.3 degrees indicated that uranium was not present in the source. All of the sources tested with an angle below this 40.3 degree cutoff contained some type of uranium. To determine whether the uranium was processed or naturally occurring, 18.0 degrees was chosen as the upper limit for processed uranium sources. Sources that produced an angle above 18.0 degrees and below 40.3 degrees were categorized as naturally occurring uranium. The vector representation technique was able to classify the uranium sources in all of the geometries except for the geometries that included the centimeter of lead.

Studies on the Composition of the Solids in the Fumes Released in the Calciometric Process for Uranium Production; ESTUDIOS SOBRE LA COMPOSICION DE LOS SOLIDOS EN LOS HUMOS DESPRENDIDOS EN EL PROCESO CALCIOTERMICO DE OBTENCION DE URANIO

DOE Office of Scientific and Technical Information (OSTI.GOV)

Travesi, A.; de la Cruz, F.; Cellini, R.F.

1958-01-01

In the dust produced during the thermal reduction of uranium tetrafluoride with calcium, the existence of a beta activity which decays with time was confirmed. The activity was assigned to Th/sup 234/. An enrichment in Th/sup 234/ of approximately 68% over the equilibrium value was found in the dust. An anomalous enrichment in the dust was found for the trace elements Zn, Sn, Pb, and Cu. No enrichment was detected for Fe and Ni. (tr-auth)

THE QUESTIONS OF HEALTH HAZARDS FROM THE INHALATION OF INSOLUBLE URANIUM AND THORIUM OXIDES

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hodge, H.C.; Thomas, R.G.

1958-10-31

The insoluble compounds of uranium and thorium, particularly the oxides, are important in the development of atomic energy. Thc questions of health hazards from exposures to dusts of these insoluble compounds are strikily simlar in many but not all respects, Among the similarities may be listed the following facts: The insoluble compounds present no chemical hazard. Both uranium and thorium dioxides, for example, are remarkably inert physiologically. No radiation injuries have so far been described in the lungs of experimental animals inhaling dust concentrations many times the recommended MAC. The lungs of a few dogs studied seven years after excessivemore » inhalation exposures to ThO/sub 2/ gave negative histological findings although high concentrations of thorium were present. The MACs for insoluble uranium and for inxoluble thorium dusts are identical, specifically 3 x 10/sup -11/ c/1. Calculated on a radiation basis, a lower MAC is appropriate for thorium. Based on a considerable body of information from cted. For both uranium and thorium dioxides fecal excretion reflects the immediate exposure to dusty atmospheres. Urine analyses are a prime index of uranium exposure whereas the presence of the much less soluble thorium dioxide in the lung cannot be thus assessed. Breath thoron extimnations or possibly measurements using a whole body counter have been recommended as indices of thorium exposure. The fundamental question depends on the radiosensitivity of the lung and of the pulmonary lymph nodes; neither the production of radiation injury nor the production of cancer are evaluated at present with respect to dosage of radiation. The lung tissues of the dogs described above must have received several thousand rem during the 7 year period. The pulmonary lymph modes must have received considerably more radiation because the concentrations in these nodes e use of the insoluble oxides and the low MACs combine to raise recurring questions of health hazards. (auth)« less

76 FR 29240 - Environmental Impacts Statements; Notice of Availability

Federal Register 2010, 2011, 2012, 2013, 2014

2011-05-20

...-283-7681. EIS No. 20110150, Final EIS, DOE, ID, ADOPTION--Areva Eagle Rock Enrichment Facility... Uranium Enrichment Facility, Construction, Operation, and Decommission, License Issuance, Piketon, OH...

77 FR 13367 - General Electric-Hitachi Global Laser Enrichment, LLC, Proposed Laser-Based Uranium Enrichment...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-03-06

... NUCLEAR REGULATORY COMMISSION [NRC-2009-0157] General Electric-Hitachi Global Laser Enrichment... Impact Statement (EIS) for the proposed General Electric- Hitachi Global Laser Enrichment, LLC (GLE... issue a license to GLE, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) parts 30, 40...

Microbial links between sulfate reduction and metal retention in uranium- and heavy metal-contaminated soil.

PubMed

Sitte, Jana; Akob, Denise M; Kaufmann, Christian; Finster, Kai; Banerjee, Dipanjan; Burkhardt, Eva-Maria; Kostka, Joel E; Scheinost, Andreas C; Büchel, Georg; Küsel, Kirsten

2010-05-01

Sulfate-reducing bacteria (SRB) can affect metal mobility either directly by reductive transformation of metal ions, e.g., uranium, into their insoluble forms or indirectly by formation of metal sulfides. This study evaluated in situ and biostimulated activity of SRB in groundwater-influenced soils from a creek bank contaminated with heavy metals and radionuclides within the former uranium mining district of Ronneburg, Germany. In situ activity of SRB, measured by the (35)SO(4)(2-) radiotracer method, was restricted to reduced soil horizons with rates of < or =142 +/- 20 nmol cm(-3) day(-1). Concentrations of heavy metals were enriched in the solid phase of the reduced horizons, whereas pore water concentrations were low. X-ray absorption near-edge structure (XANES) measurements demonstrated that approximately 80% of uranium was present as reduced uranium but appeared to occur as a sorbed complex. Soil-based dsrAB clone libraries were dominated by sequences affiliated with members of the Desulfobacterales but also the Desulfovibrionales, Syntrophobacteraceae, and Clostridiales. [(13)C]acetate- and [(13)C]lactate-biostimulated soil microcosms were dominated by sulfate and Fe(III) reduction. These processes were associated with enrichment of SRB and Geobacteraceae; enriched SRB were closely related to organisms detected in soils by using the dsrAB marker. Concentrations of soluble nickel, cobalt, and occasionally zinc declined < or =100% during anoxic soil incubations. In contrast to results in other studies, soluble uranium increased in carbon-amended treatments, reaching < or =1,407 nM in solution. Our results suggest that (i) ongoing sulfate reduction in contaminated soil resulted in in situ metal attenuation and (ii) the fate of uranium mobility is not predictable and may lead to downstream contamination of adjacent ecosystems.

Microbial Links between Sulfate Reduction and Metal Retention in Uranium- and Heavy Metal-Contaminated Soil▿

PubMed Central

Sitte, Jana; Akob, Denise M.; Kaufmann, Christian; Finster, Kai; Banerjee, Dipanjan; Burkhardt, Eva-Maria; Kostka, Joel E.; Scheinost, Andreas C.; Büchel, Georg; Küsel, Kirsten

2010-01-01

Sulfate-reducing bacteria (SRB) can affect metal mobility either directly by reductive transformation of metal ions, e.g., uranium, into their insoluble forms or indirectly by formation of metal sulfides. This study evaluated in situ and biostimulated activity of SRB in groundwater-influenced soils from a creek bank contaminated with heavy metals and radionuclides within the former uranium mining district of Ronneburg, Germany. In situ activity of SRB, measured by the 35SO42− radiotracer method, was restricted to reduced soil horizons with rates of ≤142 ± 20 nmol cm−3 day−1. Concentrations of heavy metals were enriched in the solid phase of the reduced horizons, whereas pore water concentrations were low. X-ray absorption near-edge structure (XANES) measurements demonstrated that ∼80% of uranium was present as reduced uranium but appeared to occur as a sorbed complex. Soil-based dsrAB clone libraries were dominated by sequences affiliated with members of the Desulfobacterales but also the Desulfovibrionales, Syntrophobacteraceae, and Clostridiales. [13C]acetate- and [13C]lactate-biostimulated soil microcosms were dominated by sulfate and Fe(III) reduction. These processes were associated with enrichment of SRB and Geobacteraceae; enriched SRB were closely related to organisms detected in soils by using the dsrAB marker. Concentrations of soluble nickel, cobalt, and occasionally zinc declined ≤100% during anoxic soil incubations. In contrast to results in other studies, soluble uranium increased in carbon-amended treatments, reaching ≤1,407 nM in solution. Our results suggest that (i) ongoing sulfate reduction in contaminated soil resulted in in situ metal attenuation and (ii) the fate of uranium mobility is not predictable and may lead to downstream contamination of adjacent ecosystems. PMID:20363796

10 CFR 766.3 - Definitions.

Code of Federal Regulations, 2010 CFR

2010-01-01

... Enrichment Services System, which is the database that tracks uranium enrichment services transactions of the... invoicing and historical tracking of SWU deliveries. Use and burnup charges mean lease charges for the...

Kr ion irradiation study of the depleted-uranium alloys

NASA Astrophysics Data System (ADS)

Gan, J.; Keiser, D. D.; Miller, B. D.; Kirk, M. A.; Rest, J.; Allen, T. R.; Wachs, D. M.

2010-12-01

Fuel development for the reduced enrichment research and test reactor (RERTR) program is tasked with the development of new low enrichment uranium nuclear fuels that can be employed to replace existing high enrichment uranium fuels currently used in some research reactors throughout the world. For dispersion type fuels, radiation stability of the fuel-cladding interaction product has a strong impact on fuel performance. Three depleted-uranium alloys are cast for the radiation stability studies of the fuel-cladding interaction product using Kr ion irradiation to investigate radiation damage from fission products. SEM analysis indicates the presence of the phases of interest: U(Al, Si) 3, (U, Mo)(Al, Si) 3, UMo 2Al 20, U 6Mo 4Al 43 and UAl 4. Irradiations of TEM disc samples were conducted with 500 keV Kr ions at 200 °C to ion doses up to 2.5 × 10 19 ions/m 2 (˜10 dpa) with an Kr ion flux of 10 16 ions/m 2/s (˜4.0 × 10 -3 dpa/s). Microstructural evolution of the phases relevant to fuel-cladding interaction products was investigated using transmission electron microscopy.

77 FR 39899 - Technical Corrections

Federal Register 2010, 2011, 2012, 2013, 2014

2012-07-06

..., Nuclear material, Oil and gas exploration--well logging, Reporting and recordkeeping requirements... recordkeeping requirements, Source material, Uranium. 10 CFR Part 50 Antitrust, Classified information, Criminal... measures, Special nuclear material, Uranium enrichment by gaseous diffusion. 10 CFR Part 81 Administrative...

Hybrid interferometric/dispersive atomic spectroscopy of laser-induced uranium plasma

DOE PAGES

Morgan, Phyllis K.; Scott, Jill R.; Jovanovic, Igor

2015-12-19

An established optical emission spectroscopy technique, laser-induced breakdown spectroscopy (LIBS), holds promise for detection and rapid analysis of elements relevant for nuclear safeguards, nonproliferation, and nuclear power, including the measurement of isotope ratios. One such important application of LIBS is the measurement of uranium enrichment ( 235U/ 238U), which requires high spectral resolution (e.g., 25 pm for the 424.4 nm U II line). High-resolution dispersive spectrometers necessary for such measurements are typically bulky and expensive. We demonstrate the use of an alternative measurement approach, which is based on an inexpensive and compact Fabry–Perot etalon integrated with a low to moderatemore » resolution Czerny–Turner spectrometer, to achieve the resolution needed for isotope selectivity of LIBS of uranium in ambient air. Furthermore, spectral line widths of ~ 10 pm have been measured at a center wavelength 424.437 nm, clearly discriminating the natural from the highly enriched uranium.« less

Processing of LEU targets for {sup 99}Mo production--testing and modification of the Cintichem process

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wu, D.; Landsberger, S.; Buchholz, B.

1995-09-01

Recent experimental results on testing and modification of the Cintichem process to allow substitution of low enriched uranium (LEU) for high enriched uranium (HEU) targets are presented in this report. The main focus is on {sup 99}Mo recovery and purification by its precipitation with {alpha}-benzoin oxime. Parameters that were studied include concentrations of nitric and sulfuric acids, partial neutralization of the acids, molybdenum and uranium concentrations, and the ratio of {alpha}-benzoin oxime to molybdenum. Decontamination factors for uranium, neptunium, and various fission products were measured. Experiments with tracer levels of irradiated LEU were conducted for testing the {sup 99}Mo recoverymore » and purification during each step of the Cintichem process. Improving the process with additional processing steps was also attempted. The results indicate that the conversion of molybdenum chemical processing from HEU to LEU targets is possible.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hyder, M L; Perkins, W C; Thompson, M C

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

PREPARATION OF URANIUM TRIOXIDE

DOEpatents

Buckingham, J.S.

1959-09-01

The production of uranium trioxide from aqueous solutions of uranyl nitrate is discussed. The uranium trioxide is produced by adding sulfur or a sulfur-containing compound, such as thiourea, sulfamic acid, sulfuric acid, and ammonium sulfate, to the uranyl solution in an amount of about 0.5% by weight of the uranyl nitrate hexahydrate, evaporating the solution to dryness, and calcining the dry residue. The trioxide obtained by this method furnished a dioxide with a considerably higher reactivity with hydrogen fluoride than a trioxide prepared without the sulfur additive.

Utilization of non-weapons-grade plutonium and highly enriched uranium with breeding of the {sup 233}U isotope in the VVER reactors using thorium and heavy water

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marshalkin, V. E., E-mail: marshalkin@vniief.ru; Povyshev, V. M.

A method for joint utilization of non-weapons-grade plutonium and highly enriched uranium in the thorium–uranium—plutonium oxide fuel of a water-moderated reactor with a varying water composition (D{sub 2}O, H{sub 2}O) is proposed. The method is characterized by efficient breeding of the {sup 233}U isotope and safe reactor operation and is comparatively simple to implement.

Uranium Bio-accumulation and Cycling as revealed by Uranium Isotopes in Naturally Reduced Sediments from the Upper Colorado River Basin

NASA Astrophysics Data System (ADS)

Lefebvre, Pierre; Noël, Vincent; Jemison, Noah; Weaver, Karrie; Bargar, John; Maher, Kate

2016-04-01

Uranium (U) groundwater contamination following oxidized U(VI) releases from weathering of mine tailings is a major concern at numerous sites across the Upper Colorado River Basin (CRB), USA. Uranium(IV)-bearing solids accumulated within naturally reduced zones (NRZs) characterized by elevated organic carbon and iron sulfide compounds. Subsequent re-oxidation of U(IV)solid to U(VI)aqueous then controls the release to groundwater and surface water, resulting in plume persistence and raising public health concerns. Thus, understanding the extent of uranium oxidation and reduction within NRZs is critical for assessing the persistence of the groundwater contamination. In this study, we measured solid-phase uranium isotope fractionation (δ238/235U) of sedimentary core samples from four study sites (Shiprock, NM, Grand Junction, Rifle and Naturita, CO) using a multi-collector inductively coupled plasma mass spectrometer (MC-ICP-MS). We observe a strong correlation between U accumulation and the extent of isotopic fractionation, with Δ238U up to +1.8 ‰ between uranium-enriched and low concentration zones. The enrichment in the heavy isotopes within the NRZs appears to be especially important in the vadose zone, which is subject to variations in water table depth. According to previous studies, this isotopic signature is consistent with biotic reduction processes associated with metal-reducing bacteria. Positive correlations between the amount of iron sulfides and the accumulation of reduced uranium underline the importance of sulfate-reducing conditions for U(IV) retention. Furthermore, the positive fractionation associated with U reduction observed across all sites despite some variations in magnitude due to site characteristics, shows a regional trend across the Colorado River Basin. The maximum extent of 238U enrichment observed in the NRZ proximal to the water table further suggests that the redox cycling of uranium, with net release of U(VI) to the groundwater by non-fractionating oxidation, is occurring within this zone. Thus, release of uranium from the NRZs may play a critical role in the persistence of groundwater contamination at these sites.

The Effect of U-234 Content on the Neutronic Behavior of Uranium Systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Busch, Robert D.; Bledsoe, Keith C

2011-01-01

When analyzing uranium systems, the usual rule of thumb is to ignore the U-234 by assuming that it behaves neutronically like U-238. Thus for uranium systems, the uranium is evaluated as U-235 with everything else being U-238. The absorption cross section of U-234 is indeed qualitatively very similar to that of U-238. However, thermal absorption cross section of U-234 is about 100 times that of U-238. At low U-235 enrichments, the amount of U-234 is quite small so the impact of assuming it is U-238 is minimal. However, at high enrichments, the relative ratio of U-234 to U-238 is quitemore » large (maybe as much as 1 to 5). Thus, one would expect that some effect of using the rule of thumb might be seen in higher enriched systems. Analyses were performed on three uranium systems from the set of Benchmarks [1]. Although the benchmarks are adequately characterized as to the U-234 content, often, materials used in processing are not as well characterized. This issue may become more important with the advent of laser enrichment processes, which have little or no effect on the U-234 content. Analytical results based on the relationship of U-234 activity to that of U-235 have shown good predictive capability but with large variability in the uncertainties [2]. Rucker and Johnson noted that the actual isotopics vary with enrichment, design of the enrichment cascade, composition of the feed material, and on blending of enrichments so there is considerable uncertainty in the use of models to determine isotopics. Thus, it is important for criticality personnel to understand the effects of variation of U-234 content in fissile systems and the impact of different modeling assumptions in handling the U-234. Analyses were done on LEU, IEU and HEU benchmarks from the International Handbook. These indicate that the effect of ignoring U-234 in HEU metal systems is non-conservative while it seems to be conservative for HEU solution systems. The magnitude of change in k-effective was as high as 0.4%, which has implications on selection of administrative margins and the determination of the upper subcriticality limit.« less

Th-230 - U-238 series disequilibrium of the Olkaria rhyolites Gregory Rift Valley, Kenya: Petrogenesis

NASA Technical Reports Server (NTRS)

Black, S.; Macdonald, R.; Kelly, M.

1993-01-01

Positive correlations of (U-238/Th-230) versus Th show the rhyolites to be products of partial melting. Positive correlations of U and Cl and U and F show that the U enrichment in the rhyolites is associated with the halogen contents which may be related to the minor phenocryst phase fractionation. Instantaneous Th/U ratios exceed time integrated Th/U ratios providing further evidence of the hydrous nature of the Olkaria rhyolite source. Excess (U-238/Th-230) in the subduction related rocks has been associated to the preferential incorporation of uranium in slab derived fluids, but no evaluation of the size of this flux has been made. The majority of the Naivasha samples show a (U-238/Th-230) less than 1 and plot close to the subduction related samples indicating the Naivasha rhyolites may also have been influenced by fluids during their formation. In general samples with high (U-238/Th-230) ratios reflecting recent enrichment of uranium relative to thorium have high thorium contents, thereby the high (U-238/Th-230) ratios are restricted to the most incompatible element enriched magmas and, hence, are a good indication that the rhyolites were formed by partial melting. If a fluid phase had some influence on the formation of the rhyolites then the uranium and thorium may have some correlation with F and Cl contents which can be mirrored by the peralkalinity. Plots of uranium against F and Cl contents are shown. The positive correlation indicates that the uranium enrichments are associated with the halogen contents. There seems to be a greater correlation for U against Cl than F indicating that the U may be transported preferentially as Cl complexes.

77 FR 14838 - General Electric-Hitachi Global Laser Enrichment LLC, Commercial Laser-Based Uranium Enrichment...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-03-13

... safety, chemical process safety, fire safety, emergency management, environmental protection... the transportation of SNM of low strategic significance, human factors engineering, and electrical...

Modeling the UO2 ex-AUC pellet process and predicting the fuel rod temperature distribution under steady-state operating condition

NASA Astrophysics Data System (ADS)

Hung, Nguyen Trong; Thuan, Le Ba; Thanh, Tran Chi; Nhuan, Hoang; Khoai, Do Van; Tung, Nguyen Van; Lee, Jin-Young; Jyothi, Rajesh Kumar

2018-06-01

Modeling uranium dioxide pellet process from ammonium uranyl carbonate - derived uranium dioxide powder (UO2 ex-AUC powder) and predicting fuel rod temperature distribution were reported in the paper. Response surface methodology (RSM) and FRAPCON-4.0 code were used to model the process and to predict the fuel rod temperature under steady-state operating condition. Fuel rod design of AP-1000 designed by Westinghouse Electric Corporation, in these the pellet fabrication parameters are from the study, were input data for the code. The predictive data were suggested the relationship between the fabrication parameters of UO2 pellets and their temperature image in nuclear reactor.

Compositional changes at the interface between thorium-doped uranium dioxide and zirconium due to high-temperature annealing

NASA Astrophysics Data System (ADS)

Youn, Young-Sang; Lee, Jeongmook; Kim, Jandee; Kim, Jong-Yun

2018-06-01

Compositional changes at the interface between thorium-doped uranium dioxide (U0.97Th0.03O2) and Zr before and after annealing at 1700 °C for 18 h were studied by X-ray photoelectron spectroscopy, X-ray diffraction, and Raman spectroscopy. At room temperature, the U0.97Th0.03O2 pellet consisted of hyperstoichiometric UO2+x with UO2 and ThO2, and the Zr sample contained Zr with ZrO2. After annealing, the former contained stoichiometric UO2 with ThO2 and the latter consisted of ZrO2 along with ZrO2·2H2O.

Microstructure of RERTR DU-Alloys Irradiated with Krypton Ions

DOE Office of Scientific and Technical Information (OSTI.GOV)

J. Gan; D. Keiser; D. Wachs

2009-11-01

Fuel development for reduced enrichment research and test reactor (RERTR) program is tasked with the development of new low enrichment uranium fuels that can be employed to replace existing high enrichment uranium fuels currently used in many research and test reactors worldwide. Radiation stability of the interaction product formed at fuel-matrix interface has a strong impact on fuel performance. Three depleted uranium alloys are cast that consist of the following 5 phases of interest to be investigated: U(Si,Al)3, (U,Mo)(Si,Al)3, UMo2Al20, U6Mo4Al43 and UAl4. Irradiation of TEM disc samples with 500 keV Kr ions at 200?C to high doses up tomore » ~100 dpa were conducted using an intermediate voltage electron microscope equipped with an ion accelerator. The irradiated microstructure of the 5 phases is characterized using transmission electron microscopy. The results will be presented and the implication of the observed irradiated microstructure on the fuel performance will be discussed.« less

Fractionation of whey protein isolate with supercritical carbon dioxide to produce enriched alpha-lactalbumin and beta-lactoglobulin food ingredients

USDA-ARS?s Scientific Manuscript database

A potentially economical and environmentally friendly whey protein fractionation process was developed using supercritical carbon dioxide (SCO2) as an acid to produce enriched fractions of alpha-lactalbumin (a-LA) and beta-lactoglobulin (b-LG) from whey protein isolate. To prepare the fractions, so...

Theoretical Model for Volume Fraction of UC, 235U Enrichment, and Effective Density of Final U 10Mo Alloy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Devaraj, Arun; Prabhakaran, Ramprashad; Joshi, Vineet V.

2016-04-12

The purpose of this document is to provide a theoretical framework for (1) estimating uranium carbide (UC) volume fraction in a final alloy of uranium with 10 weight percent molybdenum (U-10Mo) as a function of final alloy carbon concentration, and (2) estimating effective 235U enrichment in the U-10Mo matrix after accounting for loss of 235U in forming UC. This report will also serve as a theoretical baseline for effective density of as-cast low-enriched U-10Mo alloy. Therefore, this report will serve as the baseline for quality control of final alloy carbon content

THE FINAL DEMISE OF EAST TENNESSEE TECHNOLOGY PARK BUILDING K-33 Health Physics Society Annual Meeting West Palm Beach, Florida June 27, 2011

DOE Office of Scientific and Technical Information (OSTI.GOV)

David A. King

2011-06-27

Building K-33 was constructed in 1954 as the final section of the five-stage uranium enrichment cascade at the Oak Ridge Gaseous Diffusion Plant (ORGDP). The two original building (K-25 and K-27) were used to produce weapons grade highly enriched uranium (HEU). Building K-29, K-31, and K-33 were added to produce low enriched uranium (LEU) for nuclear power plant fuel. During ORGDP operations K-33 produced a peak enrichment of 2.5%. Thousands of tons of reactor tails fed into gaseous diffusion plants in the 1950s and early 1960s introducing some fission products and transuranics. Building K-33 was a two-story, 25-meters (82-feet) tallmore » structure with approximately 30 hectare (64 acres) of floor space. The Operations (first) Floor contained offices, change houses, feed vaporization rooms, and auxiliary equipment to support enrichment operations. The Cell (second) Floor contained the enrichment process equipment and was divided into eight process units (designated K-902-1 through K-902-8). Each unit contained ten cells, and each cell contained eight process stages (diffusers) for a total of 640 enrichment stages. 1985: LEU buildings were taken off-line after the anticipated demand for uranium enrichment failed to materialize. 1987: LEU buildings were placed in permanent shutdown. Process equipment were maintained in a shutdown state. 1997: DOE signed an Action Memorandum for equipment removal and decontamination of Buildings K-29, K-31, K-33; BNFL awarded contract to reindustrialize the buildings under the Three Buildings D&D and Recycle Project. 2002: Equipment removal complete and effort shifts to vacuuming, chemical cleaning, scabbling, etc. 2005: Decontamination efforts in K-33 cease. Building left with significant {sup 99}Tc contamination on metal structures and PCB contamination in concrete. Uranium, transuranics, and fission products also present on building shell. 2009: DOE targets Building K-33 for demolition. 2010: ORAU contracted to characterize Building K-33 for final disposition at the Environmental Management Waste Management Facility (EMWMF) in Oak Ridge. ORAU collected 439 samples from May and June. LATA Sharp started removing transite panels in September. 2011: LATA Sharp began demolition in January and expects the last waste shipment to EMWMF in September. Approximately 237,000 m{sup 3} (310,000 yd{sup 3}, bulked) of waste taken to EMWMF in 23,000 truckloads expected by project completion.« less

77 FR 14360 - Environmental Impacts Statements; Notice of Availability

Federal Register 2010, 2011, 2012, 2013, 2014

2012-03-09

... Global Laser Enrichment LLC Facility, Issuance of License to Construct, Operate, and Decommission a Laser-Based Uranium Enrichment Facility, Wilmington, NC, Review Period Ends: 04/09/2012, Contact: Jennifer A...

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Daily, Charles R.

2015-10-01

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

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

K. J. Allen; I. Bolshinsky; L. L. Biro

2010-03-01

In June 2009 two air shipments transported both unirradiated (fresh) and irradiated (spent) Russian-origin highly enriched uranium (HEU) nuclear fuel from two research reactors in Romania to the Russian Federation 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 Returnmore » 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.« less

Technical solutions to nonproliferation challenges

NASA Astrophysics Data System (ADS)

Satkowiak, Lawrence

2014-05-01

The threat of nuclear terrorism is real and poses a significant challenge to both U.S. and global security. For terrorists, the challenge is not so much the actual design of an improvised nuclear device (IND) but more the acquisition of the special nuclear material (SNM), either highly enriched uranium (HEU) or plutonium, to make the fission weapon. This paper provides two examples of technical solutions that were developed in support of the nonproliferation objective of reducing the opportunity for acquisition of HEU. The first example reviews technologies used to monitor centrifuge enrichment plants to determine if there is any diversion of uranium materials or misuse of facilities to produce undeclared product. The discussion begins with a brief overview of the basics of uranium processing and enrichment. The role of the International Atomic Energy Agency (IAEA), its safeguard objectives and how the technology evolved to meet those objectives will be described. The second example focuses on technologies developed and deployed to monitor the blend down of 500 metric tons of HEU from Russia's dismantled nuclear weapons to reactor fuel or low enriched uranium (LEU) under the U.S.-Russia HEU Purchase Agreement. This reactor fuel was then purchased by U.S. fuel fabricators and provided about half the fuel for the domestic power reactors. The Department of Energy established the HEU Transparency Program to provide confidence that weapons usable HEU was being blended down and thus removed from any potential theft scenario. Two measurement technologies, an enrichment meter and a flow monitor, were combined into an automated blend down monitoring system (BDMS) and were deployed to four sites in Russia to provide 24/7 monitoring of the blend down. Data was downloaded and analyzed periodically by inspectors to provide the assurances required.

Technical solutions to nonproliferation challenges

DOE Office of Scientific and Technical Information (OSTI.GOV)

Satkowiak, Lawrence

2014-05-09

The threat of nuclear terrorism is real and poses a significant challenge to both U.S. and global security. For terrorists, the challenge is not so much the actual design of an improvised nuclear device (IND) but more the acquisition of the special nuclear material (SNM), either highly enriched uranium (HEU) or plutonium, to make the fission weapon. This paper provides two examples of technical solutions that were developed in support of the nonproliferation objective of reducing the opportunity for acquisition of HEU. The first example reviews technologies used to monitor centrifuge enrichment plants to determine if there is any diversionmore » of uranium materials or misuse of facilities to produce undeclared product. The discussion begins with a brief overview of the basics of uranium processing and enrichment. The role of the International Atomic Energy Agency (IAEA), its safeguard objectives and how the technology evolved to meet those objectives will be described. The second example focuses on technologies developed and deployed to monitor the blend down of 500 metric tons of HEU from Russia's dismantled nuclear weapons to reactor fuel or low enriched uranium (LEU) under the U.S.-Russia HEU Purchase Agreement. This reactor fuel was then purchased by U.S. fuel fabricators and provided about half the fuel for the domestic power reactors. The Department of Energy established the HEU Transparency Program to provide confidence that weapons usable HEU was being blended down and thus removed from any potential theft scenario. Two measurement technologies, an enrichment meter and a flow monitor, were combined into an automated blend down monitoring system (BDMS) and were deployed to four sites in Russia to provide 24/7 monitoring of the blend down. Data was downloaded and analyzed periodically by inspectors to provide the assurances required.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Burkes, Douglas E.; Senor, David J.; Casella, Andrew M.

Numerous global programs are focused on the continued development of existing and new research and test reactor fuels to achieve maximum attainable uranium loadings to support the conversion of a number of the world’s remaining high-enriched uranium fueled reactors to low-enriched uranium fuel. Some of these programs are focused on development and qualification of a fuel design that consists of a uranium-molybdenum (U-Mo) alloy dispersed in an aluminum matrix as one option for reactor conversion. The current paper extends a failure model originally developed for UO2-stainless steel dispersion fuels and used currently available thermal-mechanical property information for the materials ofmore » interest in the current proposed design. A number of fabrication and irradiation parameters were investigated to understand the conditions at which failure of the matrix, classified as pore formation in the matrix, might occur. The results compared well with experimental observations published as part of the Reduced Enrichment for Research and Test Reactors (RERTR)-6 and -7 mini-plate experiments. Fission rate, a function of the 235U enrichment, appeared to be the most influential parameter in premature failure, mainly as a result of increased interaction layer formation and operational temperature, which coincidentally decreased the yield strength of the matrix and caused more rapid fission gas production and recoil into the surrounding matrix material. Addition of silicon to the matrix appeared effective at reducing the rate of interaction layer formation and can extend the performance of a fuel plate under a certain set of irradiation conditions, primarily moderate heat flux and burnup. Increasing the dispersed fuel particle diameter may also be effective, but only when combined with other parameters, e.g., lower enrichment and increased Si concentration. The model may serve as a valuable tool in initial experimental design.« less

Gas centrifuge enrichment plants inspection frequency and remote monitoring issues for advanced safeguards implementation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Boyer, Brian David; Erpenbeck, Heather H; Miller, Karen A

2010-09-13

Current safeguards approaches used by the IAEA at gas centrifuge enrichment plants (GCEPs) need enhancement in order to verify declared low enriched uranium (LEU) production, detect undeclared LEU production and detect high enriched uranium (BEU) production with adequate probability using non destructive assay (NDA) techniques. At present inspectors use attended systems, systems needing the presence of an inspector for operation, during inspections to verify the mass and {sup 235}U enrichment of declared cylinders of uranium hexafluoride that are used in the process of enrichment at GCEPs. This paper contains an analysis of how possible improvements in unattended and attended NDAmore » systems including process monitoring and possible on-site destructive analysis (DA) of samples could reduce the uncertainty of the inspector's measurements providing more effective and efficient IAEA GCEPs safeguards. We have also studied a few advanced safeguards systems that could be assembled for unattended operation and the level of performance needed from these systems to provide more effective safeguards. The analysis also considers how short notice random inspections, unannounced inspections (UIs), and the concept of information-driven inspections can affect probability of detection of the diversion of nuclear material when coupled to new GCEPs safeguards regimes augmented with unattended systems. We also explore the effects of system failures and operator tampering on meeting safeguards goals for quantity and timeliness and the measures needed to recover from such failures and anomalies.« less

Method for fluorination of uranium oxide

DOEpatents

Petit, George S.

1987-01-01

Highly pure uranium hexafluoride is made from uranium oxide and fluorine. The uranium oxide, which includes UO.sub.3, UO.sub.2, U.sub.3 O.sub.8 and mixtures thereof, is introduced together with a small amount of a fluorine-reactive substance, selected from alkali chlorides, silicon dioxide, silicic acid, ferric oxide, and bromine, into a constant volume reaction zone. Sufficient fluorine is charged into the zone at a temperature below approximately 0.degree. C. to provide an initial pressure of at least approximately 600 lbs/sq. in. at the ambient atmospheric temperature. The temperature is then allowed to rise in the reaction zone until reaction occurs.

A method for phenomenological and chemical kinetics study of autocatalytic reactive dissolution by optical microscopy. The case of uranium dioxide dissolution in nitric acid media

NASA Astrophysics Data System (ADS)

Marc, Philippe; Magnaldo, Alastair; Godard, Jérémy; Schaer, Éric

2018-03-01

Dissolution is a milestone of the head-end of hydrometallurgical processes, as the stabilization rates of the chemical elements determine the process performance and hold-up. This study aims at better understanding the chemical and physico-chemical phenomena of uranium dioxide dissolution reactions in nitric acid media in the Purex process, which separates the reusable materials and the final wastes of the spent nuclear fuels. It has been documented that the attack of sintering-manufactured uranium dioxide solids occurs through preferential attack sites, which leads to the development of cracks in the solids. Optical microscopy observations show that in some cases, the development of these cracks leads to the solid cleavage. It is shown here that the dissolution of the detached fragments is much slower than the process of the complete cleavage of the solid, and occurs with no disturbing phenomena, like gas bubbling. This fact has motivated the measurement of dissolution kinetics using optical microscopy and image processing. By further discriminating between external resistance and chemical reaction, the "true" chemical kinetics of the reaction have been measured, and the highly autocatalytic nature of the reaction confirmed. Based on these results, the constants of the chemical reactions kinetic laws have also been evaluated.

75 FR 36447 - Notice of Availability of Draft Environmental Impact Statement and Public Meetings for the...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-06-25

... Statement and Public Meetings for the General Electric-Hitachi Global Laser Enrichment, LLC Proposed Laser... the proposed General Electric-Hitachi (GEH) Global Laser Enrichment (GLE) Uranium Enrichment Facility... to locate the facility on the existing General Electric Company (GE) site near Wilmington, North...

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

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-24

... support safe operation of Autoclave 2 of the facility have been constructed in accordance with the... Inspection Reports Regarding Louisiana Energy Services, National Enrichment Facility, Eunice, New Mexico... Louisiana Energy Services (LES), LLC, National Enrichment Facility in Eunice, New Mexico, and has authorized...

Hunting a Black Swan: Policy Options for America’s Police in Preventing Radiological/Nuclear Terrorism

DTIC Science & Technology

2012-09-01

patrol vehicles. The Department’s Counter-Terror Operations Unit serves as the program coordinator and as the archetypical NIMS Type I Team. The...is defined by Title I of the Atomic Energy Act of 1954 as plutonium, uranium-233, or uranium enriched in the isotopes uranium-233 or uranium...end of World War II. Radioactive Materials—materials that contain radioactive atoms . Radioactive atoms are unstable; that is, they have too much

METHOD OF MAKING UO$sub 2$-Bi SLURRIES

DOEpatents

Hahn, H.T.

1960-05-24

A process is given of preparing an easily dispersible slurry of uranium dioxide in bismuth. A mixture of bismuth oxide, uranium, and bismuth are heated in a capsule to a temperature over the melting point of bismuth oxide. The amount of bismuth oxide used is less than that stoichiometrically required because the oxygen in the capsule also enters into the reaction.

Nuclear Fuel Reprocessing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harold F. McFarlane; Terry Todd

2013-11-01

Reprocessing is essential to closing nuclear fuel cycle. Natural uranium contains only 0.7 percent 235U, the fissile (see glossary for technical terms) isotope that produces most of the fission energy in a nuclear power plant. Prior to being used in commercial nuclear fuel, uranium is typically enriched to 3–5% in 235U. If the enrichment process discards depleted uranium at 0.2 percent 235U, it takes more than seven tonnes of uranium feed to produce one tonne of 4%-enriched uranium. Nuclear fuel discharged at the end of its economic lifetime contains less one percent 235U, but still more than the natural ore.more » Less than one percent of the uranium that enters the fuel cycle is actually used in a single pass through the reactor. The other naturally occurring isotope, 238U, directly contributes in a minor way to power generation. However, its main role is to transmute into plutoniumby neutron capture and subsequent radioactive decay of unstable uraniumand neptuniumisotopes. 239Pu and 241Pu are fissile isotopes that produce more than 40% of the fission energy in commercially deployed reactors. It is recovery of the plutonium (and to a lesser extent the uranium) for use in recycled nuclear fuel that has been the primary focus of commercial reprocessing. Uraniumtargets irradiated in special purpose reactors are also reprocessed to obtain the fission product 99Mo, the parent isotope of technetium, which is widely used inmedical procedures. Among the fission products, recovery of such expensive metals as platinum and rhodium is technically achievable, but not economically viable in current market and regulatory conditions. During the past 60 years, many different techniques for reprocessing used nuclear fuel have been proposed and tested in the laboratory. However, commercial reprocessing has been implemented along a single line of aqueous solvent extraction technology called plutonium uranium reduction extraction process (PUREX). Similarly, hundreds of types of reactor fuels have been irradiated for different purposes, but the vast majority of commercial fuel is uranium oxide clad in zirconium alloy tubing. As a result, commercial reprocessing plants have relatively narrow technical requirements for used nuclear that is accepted for processing.« less

Micro-SHINE Uranyl Sulfate Irradiations at the Linac

DOE Office of Scientific and Technical Information (OSTI.GOV)

Youker, Amanda J.; Kalensky, Michael; Chemerisov, Sergey

2016-08-01

Peroxide formation due to water radiolysis in a uranyl sulfate solution is a concern for the SHINE Medical Technologies process in which Mo-99 is generated from the fission of dissolved low enriched uranium. To investigate the effects of power density and fission on peroxide formation and uranyl-peroxide precipitation, uranyl sulfate solutions were irradiated using a 50-MeV electron linac as part of the micro-SHINE experimental setup. Results are given for uranyl sulfate solutions with both high and low enriched uranium irradiated at different linac powers.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vandegrift, George; Brown, M. Alex; Jerden, James L.

2016-09-01

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

An aerosol particle containing enriched uranium encountered during routine sampling

NASA Astrophysics Data System (ADS)

Murphy, Daniel; Froyd, Karl; Evangeliou, NIkolaos; Stohl, Andreas

2017-04-01

The composition of single aerosol particles has been measured using a laser ionization mass spectrometer during the global Atmospheric Tomography mission. The measurements were targeting the background atmosphere, not radiochemical emissions. One sub-micron particle sampled at about 7 km altitude near the Aleutian Islands contained uranium with approximately 3% 235U. It is the only particle with enriched uranium out of millions of particles sampled over several decades of measurements with this instrument. The particle also contained vanadium, alkali metals, and organic material similar to that present in emissions from combustion of heavy oil. No zirconium or other metals that might be characteristic of nuclear reactors were present, probably suggesting a source other than Fukushima or Chernobyl. Back trajectories suggest several areas in Asia that might be sources for the particle.

INTERNAL EXPOSURE TO URANIUM IN A POOLED COHORT OF GASEOUS DIFFUSION PLANT WORKERS

PubMed Central

Anderson, Jeri L.; Apostoaei, A. Iulian; Yiin, James H.; Fleming, Donald A.; Tseng, Chih-Yu; Chen, Pi-Hsueh

2015-01-01

Intakes and absorbed organ doses were estimated for 29 303 workers employed at three former US gaseous diffusion plants as part of a study of cause-specific mortality and cancer incidence in uranium enrichment workers. Uranium urinalysis data (>600 000 urine samples) were available for 58 % of the pooled cohort. Facility records provided uranium gravimetric and radioactivity concentration data and allowed estimation of enrichment levels of uranium to which workers may have been exposed. Urine data were generally recorded with facility department numbers, which were also available in study subjects’ work histories. Bioassay data were imputed for study subjects with no recorded sample results (33 % of pooled cohort) by assigning department average urine uranium concentration. Gravimetric data were converted to 24-h uranium activity excretion using department average specific activities. Intakes and organ doses were calculated assuming chronic exposure by inhalation to a 5-µm activity median aerodynamic diameter aerosol of soluble uranium. Median intakes varied between 0.31 and 0.74 Bq d−1 for the three facilities. Median organ doses for the three facilities varied between 0.019 and 0.051, 0.68 and 1.8, 0.078 and 0.22, 0.28 and 0.74, and 0.094 and 0.25 mGy for lung, bone surface, red bone marrow, kidneys, and liver, respectively. Estimated intakes and organ doses for study subjects with imputed bioassay data were similar in magnitude. PMID:26113578

Liquid Thermal Diffusion during the Manhattan Project

NASA Astrophysics Data System (ADS)

Cameron Reed, B.

2011-06-01

On the basis of Manhattan Engineer District documents, a little known Naval Research Laboratory report of 1946, and other sources, I construct a more complete history of the liquid-thermal-diffusion method of uranium enrichment during World War II than is presented in official histories of the Manhattan Project. This method was developed by Philip Abelson (1913-2004) and put into operation at the rapidly-constructed S-50 plant at Oak Ridge, Tennessee, which was responsible for the first stage of uranium enrichment, from 0.72% to 0.85% U-235, producing nearly 45,000 pounds of enriched U-235 by July 1945 at a cost of just under 20 million. I review the history, design, politics, construction, and operation of the S-50 liquid-thermal-diffusion plant.

Depleted and natural uranium: chemistry and toxicological effects.

PubMed

Craft, Elena; Abu-Qare, Aquel; Flaherty, Meghan; Garofolo, Melissa; Rincavage, Heather; Abou-Donia, Mohamed

2004-01-01

Depleted uranium (DU) is a by-product from the chemical enrichment of naturally occurring uranium. Natural uranium is comprised of three radioactive isotopes: (238)U, (235)U, and (234)U. This enrichment process reduces the radioactivity of DU to roughly 30% of that of natural uranium. Nonmilitary uses of DU include counterweights in airplanes, shields against radiation in medical radiotherapy units and transport of radioactive isotopes. DU has also been used during wartime in heavy tank armor, armor-piercing bullets, and missiles, due to its desirable chemical properties coupled with its decreased radioactivity. DU weapons are used unreservedly by the armed forces. Chemically and toxicologically, DU behaves similarly to natural uranium metal. Although the effects of DU on human health are not easily discerned, they may be produced by both its chemical and radiological properties. DU can be toxic to many bodily systems, as presented in this review. Most importantly, normal functioning of the kidney, brain, liver, and heart can be affected by DU exposure. Numerous other systems can also be affected by DU exposure, and these are also reviewed. Despite the prevalence of DU usage in many applications, limited data exist regarding the toxicological consequences on human health. This review focuses on the chemistry, pharmacokinetics, and toxicological effects of depleted and natural uranium on several systems in the mammalian body. A section on risk assessment concludes the review.

Scoping study to expedite development of a field deployable and portable instrument for UF6 enrichment assay

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chan, George; Valentine, John D.; Russo, Richard E.

The primary objective of the present study is to identity the most promising, viable technologies that are likely to culminate in an expedited development of the next-generation, field-deployable instrument for providing rapid, accurate, and precise enrichment assay of uranium hexafluoride (UF6). UF6 is typically involved, and is arguably the most important uranium compound, in uranium enrichment processes. As the first line of defense against proliferation, accurate analytical techniques to determine the uranium isotopic distribution in UF6 are critical for materials verification, accounting, and safeguards at enrichment plants. As nuclear fuel cycle technology becomes more prevalent around the world, international nuclearmore » safeguards and interest in UF6 enrichment assay has been growing. At present, laboratory-based mass spectrometry (MS), which offers the highest attainable analytical accuracy and precision, is the technique of choice for the analysis of stable and long-lived isotopes. Currently, the International Atomic Energy Agency (IAEA) monitors the production of enriched UF6 at declared facilities by collecting a small amount (between 1 to 10 g) of gaseous UF6 into a sample bottle, which is then shipped under chain of custody to a central laboratory (IAEA’s Nuclear Materials Analysis Laboratory) for high-precision isotopic assay by MS. The logistics are cumbersome and new shipping regulations are making it more difficult to transport UF6. Furthermore, the analysis is costly, and results are not available for some time after sample collection. Hence, the IAEA is challenged to develop effective safeguards approaches at enrichment plants. In-field isotopic analysis of UF6 has the potential to substantially reduce the time, logistics and expense of sample handling. However, current laboratory-based MS techniques require too much infrastructure and operator expertise for field deployment and operation. As outlined in the IAEA Department of Safeguards Long-Term R&D Plan, 2012–2023, one of the IAEA long-term R&D needs is to “develop tools and techniques to enable timely, potentially real-time, detection of HEU (Highly Enriched Uranium) production in LEU (Lowly Enriched Uranium) enrichment facilities” (Milestone 5.2). Because it is common that the next generation of analytical instruments is driven by technologies that are either currently available or just now emerging, one reasonable and practical approach to project the next generation of chemical instrumentation is to track the recent trends and to extrapolate them. This study adopted a similar approach, and an extensive literature review on existing and emerging technologies for UF6 enrichment assay was performed. The competitive advantages and current limitations of different analytical techniques for in-field UF6 enrichment assay were then compared, and the main gaps between needs and capabilities for their field use were examined. Subsequently, based on these results, technologies for the next-generation field-deployable instrument for UF6 enrichment assay were recommended. The study was organized in a way that a suite of assessment metric was first identified. Criteria used in this evaluation are presented in Section 1 of this report, and the most important ones are described briefly in the next few paragraphs. Because one driving force for in-field UF6 enrichment assay is related to the demanding transportation regulation for gaseous UF6, Section 2 contains a review of solid sorbents that convert and immobilized gaseous UF6 to a solid state, which is regarded as more transportation friendly and is less regulated. Furthermore, candidate solid sorbents, which show promise in mating with existing and emerging assay technologies, also factor into technology recommendations. Extensive literature reviews on existing and emerging technologies for UF6 enrichment assay, covering their scientific principles, instrument options, and current limitations are detailed in Sections 3 and 4, respectively. In Section 5, the technological gaps as well as start-of-the-art and commercial off-the-shelf components that can be adopted to expedite the development of a fieldable or portable UF6 enrichment-assay instrument are identified and discussed. Finally, based on the results of the review, requirements and recommendations for developing the next-generation field-deployable instrument for UF6 enrichment assay are presented in Section 6.« less

Physical characterization of uranium oxide pellets and powder applied in the Nuclear Forensics International Technical Working Group Collaborative Materials Exercise 4

DOE Office of Scientific and Technical Information (OSTI.GOV)

Griffiths, Grant; Keegan, E.; Young, E.

Physical characterization is one of the most broad and important categories of techniques to apply in a nuclear forensic examination. Physical characterization techniques vary from simple weighing and dimensional measurements to complex sample preparation and scanning electron microscopy-electron backscatter diffraction analysis. This paper reports on the physical characterization conducted by several international laboratories participating in the fourth Collaborative Materials Exercise, organized by the Nuclear Forensics International Technical Working Group. Methods include a range of physical measurements, microscopy-based observations, and profilometry. In conclusion, the value of these results for addressing key investigative questions concerning two uranium dioxide pellets and a uraniummore » dioxide powder is discussed.« less

Physical characterization of uranium oxide pellets and powder applied in the Nuclear Forensics International Technical Working Group Collaborative Materials Exercise 4

DOE PAGES

Griffiths, Grant; Keegan, E.; Young, E.; ...

2018-01-06

Physical characterization is one of the most broad and important categories of techniques to apply in a nuclear forensic examination. Physical characterization techniques vary from simple weighing and dimensional measurements to complex sample preparation and scanning electron microscopy-electron backscatter diffraction analysis. This paper reports on the physical characterization conducted by several international laboratories participating in the fourth Collaborative Materials Exercise, organized by the Nuclear Forensics International Technical Working Group. Methods include a range of physical measurements, microscopy-based observations, and profilometry. In conclusion, the value of these results for addressing key investigative questions concerning two uranium dioxide pellets and a uraniummore » dioxide powder is discussed.« less

Mortality in a Combined Cohort of Uranium Enrichment Workers

PubMed Central

Yiin, James H.; Anderson, Jeri L.; Daniels, Robert D.; Bertke, Stephen J.; Fleming, Donald A.; Tollerud, David J.; Tseng, Chih-Yu; Chen, Pi-Hsueh; Waters, Kathleen M.

2017-01-01

Objective To examine the patterns of cause-specific mortality and relationship between internal exposure to uranium and specific causes in a pooled cohort of 29,303 workers employed at three former uranium enrichment facilities in the United States with follow-up through 2011. Methods Cause-specific standardized mortality ratios (SMRs) for the full cohort were calculated with the U.S. population as referent. Internal comparison of the dose-response relation between selected outcomes and estimated organ doses was evaluated using regression models. Results External comparison with the U.S. population showed significantly lower SMRs in most diseases in the pooled cohort. Internal comparison showed positive associations of absorbed organ doses with multiple myeloma, and to a lesser degree with kidney cancer. Conclusion In general, these gaseous diffusion plant workers had significantly lower SMRs than the U.S. population. The internal comparison however, showed associations between internal organ doses and diseases associated with uranium exposure in previous studies. PMID:27753121

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rathbun, R.

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

Design of thermal neutron beam based on an electron linear accelerator for BNCT.

PubMed

Zolfaghari, Mona; Sedaghatizadeh, Mahmood

2016-12-01

An electron linear accelerator (Linac) can be used for boron neutron capture therapy (BNCT) by producing thermal neutron flux. In this study, we used a Varian 2300 C/D Linac and MCNPX.2.6.0 code to simulate an electron-photoneutron source for use in BNCT. In order to decelerate the produced fast neutrons from the photoneutron source, which optimize the thermal neutron flux, a beam-shaping assembly (BSA) was simulated. After simulations, a thermal neutron flux with sharp peak at the beam exit was obtained in the order of 3.09×10 8 n/cm 2 s and 6.19×10 8 n/cm 2 s for uranium and enriched uranium (10%) as electron-photoneutron sources respectively. Also, in-phantom dose analysis indicates that the simulated thermal neutron beam can be used for treatment of shallow skin melanoma in time of about 85.4 and 43.6min for uranium and enriched uranium (10%) respectively. Copyright © 2016. Published by Elsevier Ltd.

Higher Resolution Neutron Velocity Spectrometer Measurements of Enriched Uranium

DOE R&D Accomplishments Database

Rainwater, L. J.; Havens, W. W. Jr.

1950-08-09

The slow neutron transmission of a sample of enriched U containing 3.193 gm/cm2 was investigated with a resolution width of 1 microsec/m. Results of transmission measurements are shown graphically. (B.J.H.)

Evolution of spent nuclear fuel in dry storage conditions for millennia and beyond

NASA Astrophysics Data System (ADS)

Wiss, Thierry; Hiernaut, Jean-Pol; Roudil, Danièle; Colle, Jean-Yves; Maugeri, Emilio; Talip, Zeynep; Janssen, Arne; Rondinella, Vincenzo; Konings, Rudy J. M.; Matzke, Hans-Joachim; Weber, William J.

2014-08-01

Significant amounts of spent uranium dioxide nuclear fuel are accumulating worldwide from decades of commercial nuclear power production. While such spent fuel is intended to be reprocessed or disposed in geologic repositories, out-of-reactor radiation damage from alpha decay can be detrimental to its structural stability. Here we report on an experimental study in which radiation damage in plutonium dioxide, uranium dioxide samples doped with short-lived alpha-emitters and urano-thorianite minerals have been characterized by XRD, transmission electron microscopy, thermal desorption spectrometry and hardness measurements to assess the long-term stability of spent nuclear fuel to substantial alpha-decay doses. Defect accumulation is predicted to result in swelling of the atomic structure and decrease in fracture toughness; whereas, the accumulation of helium will produce bubbles that result in much larger gaseous-induced swelling that substantially increases the stresses in the constrained spent fuel. Based on these results, the radiation-ageing of highly-aged spent nuclear fuel over more than 10,000 years is predicted.

Effect of carbon dioxide enrichment on health-promoting compounds and organoleptic properties of tomato fruits grown in greenhouse.

PubMed

Zhang, Zhiming; Liu, Lihong; Zhang, Min; Zhang, Yongsong; Wang, Qiaomei

2014-06-15

The objective of the present study was to evaluate the effect of carbon dioxide (CO2) enrichment on the main health-promoting compounds and organoleptic characteristics of tomato (Solanum lycopersicum) fruits grown in greenhouse. The contents of health-promoting compounds, including lycopene, β-carotene, and ascorbic acid, as well as the flavour, indicated by sugars, titrable acidity, and sugar/acid ratio, were markedly increased in CO2 enrichment fruits. Furthermore, CO2 enrichment significantly enhanced other organoleptic characteristics, including colour, firmness, aroma, and sensory attributes in tomato fruits. The results indicated that CO2 enrichment has potential in promoting the nutritional value and organoleptic characteristics of tomatoes. Copyright © 2013 Elsevier Ltd. All rights reserved.

Performance and Mechanism of Uranium Adsorption from Seawater to Poly(dopamine)-Inspired Sorbents.

PubMed

Wu, Fengcheng; Pu, Ning; Ye, Gang; Sun, Taoxiang; Wang, Zhe; Song, Yang; Wang, Wenqing; Huo, Xiaomei; Lu, Yuexiang; Chen, Jing

2017-04-18

Developing facile and robust technologies for effective enrichment of uranium from seawater is of great significance for resource sustainability and environmental safety. By exploiting mussel-inspired polydopamine (PDA) chemistry, diverse types of PDA-functionalized sorbents including magnetic nanoparticle (MNP), ordered mesoporous carbon (OMC), and glass fiber carpet (GFC) were synthesized. The PDA functional layers with abundant catechol and amine/imine groups provided an excellent platform for binding to uranium. Due to the distinctive structure of PDA, the sorbents exhibited multistage kinetics which was simultaneously controlled by chemisorption and intralayer diffusion. Applying the diverse PDA-modified sorbents for enrichment of low concentration (parts per billion) uranium in laboratory-prepared solutions and unpurified seawater was fully evaluated under different scenarios: that is, by batch adsorption for MNP and OMC and by selective filtration for GFC. Moreover, high-resolution X-ray photoelectron spectroscopic and extended X-ray absorption fine structure studies were performed for probing the underlying coordination mechanism between PDA and U(VI). The catechol hydroxyls of PDA were identified as the main bidentate ligands to coordinate U(VI) at the equatorial plane. This study assessed the potential of versatile PDA chemistry for development of efficient uranium sorbents and provided new insights into the interaction mechanism between PDA and uranium.

Petrochemical and Mineralogical Constraints on the Source and Processes of Uranium Mineralisation in the Granitoids of Zing-Monkin Area, Adamawa Massif, NE Nigeria

DOE Office of Scientific and Technical Information (OSTI.GOV)

Haruna, I. V., E-mail: vela_hi@yahoo.co.uk; Orazulike, D. M.; Ofulume, A. B.

Zing-Monkin area, located in the northern part of Adamawa Massif, is underlain by extensive exposures of moderately radioactive granodiorites, anatectic migmatites, equigranular granites, porphyritic granites and highly radioactive fine-grained granites with minor pegmatites. Selected major and trace element petrochemical investigations of the rocks show that a progression from granodiorite through migmatite to granites is characterised by depletion of MgO, CaO, Fe{sub 2}O{sub 3,} Sr, Ba, and Zr, and enrichment of SiO{sub 2} and Rb. This trend is associated with uranium enrichment and shows a chemical gradation from the more primitive granodiorite to the more evolved granites. Electron microprobe analysis showsmore » that the uranium is content in uranothorite and in accessories, such as monazite, titanite, apatite, epidote and zircon. Based on petrochemical and mineralogical data, the more differentiated granitoids (e.g., fine-grained granite) bordering the Benue Trough are the immediate source of the uranium prospect in Bima Sandstone within the Trough. Uranium was derived from the granitoids by weathering and erosion. Transportation and subsequent interaction with organic matter within the Bima Sandstone led to precipitation of insoluble secondary uranium minerals in the Benue Trough.« less

Oxidant K edge x-ray emission spectroscopy of UF 4 and UO 2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tobin, J. G.; Yu, S. -W.; Qiao, R.

The K-Edge (1s) x-ray emission spectroscopy of uranium tetrafluoride and uranium dioxide were compared to each other and to the results of a pair of earlier cluster calculations. Here, using a very simplified approach, it is possible to qualitatively reconstruct the main features of the x-ray emission spectra from the cluster calculation state energies and 2p percentages.

Oxidant K edge x-ray emission spectroscopy of UF 4 and UO 2

DOE PAGES

Tobin, J. G.; Yu, S. -W.; Qiao, R.; ...

2018-01-31

The K-Edge (1s) x-ray emission spectroscopy of uranium tetrafluoride and uranium dioxide were compared to each other and to the results of a pair of earlier cluster calculations. Here, using a very simplified approach, it is possible to qualitatively reconstruct the main features of the x-ray emission spectra from the cluster calculation state energies and 2p percentages.

Active interrogation of highly enriched uranium

NASA Astrophysics Data System (ADS)

Fairrow, Nannette Lea

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

Conversion Preliminary Safety Analysis Report for the NIST Research Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Diamond, D. J.; Baek, J. S.; Hanson, A. L.

The NIST Center for Neutron Research (NCNR) is a reactor-laboratory complex providing the National Institute of Standards and Technology (NIST) and the nation with a world-class facility for the performance of neutron-based research. The heart of this facility is the NIST research reactor (aka NBSR); a heavy water moderated and cooled reactor operating at 20 MW. It is fueled with high-enriched uranium (HEU) fuel elements. A Global Threat Reduction Initiative (GTRI) program is underway to convert the reactor to low-enriched uranium (LEU) fuel. This program includes the qualification of the proposed fuel, uranium and molybdenum alloy foil clad in anmore » aluminum alloy, and the development of the fabrication techniques. This report is a preliminary version of the Safety Analysis Report (SAR) that would be submitted to the U.S. Nuclear Regulatory Commission (NRC) for approval prior to conversion. The report follows the recommended format and content from the NRC codified in NUREG-1537, “Guidelines for Preparing and Reviewing Applications for the Licensing of Non-power Reactors,” Chapter 18, “Highly Enriched to Low-Enriched Uranium Conversions.” The emphasis in any conversion SAR is to explain the differences between the LEU and HEU cores and to show the acceptability of the new design; there is no need to repeat information regarding the current reactor that will not change upon conversion. Hence, as seen in the report, the bulk of the SAR is devoted to Chapter 4, Reactor Description, and Chapter 13, Safety Analysis.« less

In-situ, time resolved monitoring of uranium in BFS:OPC grout. Part 2: Corrosion in water.

PubMed

Stitt, C A; Paraskevoulakos, C; Banos, A; Harker, N J; Hallam, K R; Pullin, H; Davenport, A; Street, S; Scott, T B

2018-06-18

To reflect potential conditions in a geological disposal facility, uranium was encapsulated in grout and submersed in de-ionised water for time periods between 2-47 weeks. Synchrotron X-ray Powder Diffraction and X-ray Tomography were used to identify the dominant corrosion products and measure their dimensions. Uranium dioxide was observed as the dominant corrosion product and time dependent thickness measurements were used to calculate oxidation rates. The effectiveness of physical and chemical grout properties to uranium corrosion and mobilisation is discussed and Inductively Coupled Plasma Mass Spectrometry was used to measure 238 U (aq) content in the residual water of several samples.

Corrosion Evaluation of RERTR Uranium Molybdenum Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

A K Wertsching

2012-09-01

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

Analysis of Tank 13H (HTF-13-14-156, 157) Surface and Subsurface Supernatant Samples in Support of Enrichment Control, Corrosion Control and Sodium Aluminosilicate Formation Potential Programs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Oji, L. N.

2015-02-18

The 2H Evaporator system includes mainly Tank 43H (feed tank) and Tank 38H (drop tank) with Tank 22H acting as the DWPF recycle receipt tank. The Tank 13H is being characterized to ensure that it can be transferred to the 2H evaporator. This report provides the results of analyses on Tanks 13H surface and subsurface supernatant liquid samples to ensure compliance with the Enrichment Control Program (ECP), the Corrosion Control Program and Sodium Aluminosilicate Formation Potential in the Evaporator. The U-235 mass divided by the total uranium averaged 0.00799 (0.799 % uranium enrichment) for both the surface and subsurface Tankmore » 13H samples. This enrichment is slightly above the enrichment for Tanks 38H and 43H, where the enrichment normally ranges from 0.59 to 0.7 wt%. The U-235 concentration in Tank 13H samples ranged from 2.01E-02 to 2.63E-02 mg/L, while the U-238 concentration in Tank 13H ranged from 2.47E+00 to 3.21E+00 mg/L. Thus, the U-235/total uranium ratio is in line with the prior 2H-evaporator ECP samples. Measured sodium and silicon concentrations averaged, respectively, 2.46 M and 1.42E-04 M (3.98 mg/L) in the Tank 13H subsurface sample. The measured aluminum concentration in Tanks 13H subsurface samples averaged 2.01E-01 M.« less

Red light and carbon dioxide differentially affect growth, lipid production, and quality in the microalga, Ettlia oleoabundans.

PubMed

Yang, Ying; Weathers, Pamela

2015-01-01

Ettlia oleoabundans, a freshwater unicellular green microalga, was grown under different light qualities ± carbon dioxide-enriched air to determine the combined effects on growth and lipid production of this oleaginous species. Keeping total light intensity constant, when a portion of the cool white was replaced by red, volumetric lipid yield increased 2.8-fold mainly due to the greater yield of oleic acid, a desirable biodiesel precursor. Only 30 min of red light treatment was sufficient to increase lipid yield and quality to the same level as cultures provided red light for >14 days, indicating the potential role of red light in stimulating lipid production of this species. Carbon dioxide enrichment via air sparging enhanced exponential growth, carbon conversion efficiency, and nutrient consumption. Together, these results showed that light quality plays an important role in microalgal lipid production. Adjustment in light quality and gas delivery efficiency with carbon dioxide enrichment improved lipid yield and quality in this and possibly other oleaginous algal species.

The Feasibility of Ending HEU Fuel Use in the U.S. Navy

DOE PAGES

Philippe, Sebastian; von Hippel, Frank

2016-11-01

We report that since September 11, 2001, the U.S. government has sought to remove weapons-useable highly enriched uranium (HEU) containing 20 percent or more uranium-235 from as many locations as possible because of concerns about the possibility of nuclear terrorism.

Symposium on the reprocessing of irradiated fuels. Book 2, Session IV

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1958-12-31

Book two of this conference has a single-focused session IV entitled Nonaqueous Processing, with 8 papers. The session deals with fluoride volatility processes and pyrometallurgical or pyrochemical processes. The latter involves either an oxide drossing or molten metal extraction or fused salt extraction technique and results in only partial decontamination. Fluoride volatility processes appear to be especially favorable for recovery of enriched uranium and decontamination factors of 10/sup 7/ to 10/sup 8/ would be achieved by simpler means than those employed in solvent extraction. Data from lab research on the BrF/sub 3/ process and the ClF/sub 3/ process are givenmore » and discussed and pilot plant experience is described, all in connection with natural uranium or slightly enriched uranium processing. Fluoride volatility processes for enriched or high alloy fuels are described step by step. The economic and engineering considerations of both types of nonaqueous processing are treated separately and as fully as present knowledge allows. A comprehensive review of the chemistry of pyrometallurgical processes is included.« less

Reactor Physics Measurements and Benchmark Specifications for Oak Ridge Highly Enriched Uranium Sphere (ORSphere)

DOE PAGES

Marshall, Margaret A.

2014-11-04

In the early 1970s Dr. John T. Mihalczo (team leader), J.J. Lynn, and J.R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF) with highly enriched uranium (HEU) metal (called Oak Ridge Alloy or ORALLOY) in an effort to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s. The purpose of the Oak Ridge ORALLOY Sphere (ORSphere) experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with themore » GODIVA I experiments. Additionally, various material reactivity worths, the surface material worth coefficient, the delayed neutron fraction, the prompt neutron decay constant, relative fission density, and relative neutron importance were all measured. The critical assembly, material reactivity worths, the surface material worth coefficient, and the delayed neutron fraction were all evaluated as benchmark experiment measurements. The reactor physics measurements are the focus of this paper; although for clarity the critical assembly benchmark specifications are briefly discussed.« less

10 CFR 51.20 - Criteria for and identification of licensing and regulatory actions requiring environmental...

Code of Federal Regulations, 2010 CFR

2010-01-01

... design capacity license to operate an isotopic enrichment plant pursuant to part 50 of this chapter. (4... uranium enrichment facility. (11) Issuance of renewal of a license authorizing receipt and disposal of...

High Useful Yield and Isotopic Analysis of Uranium by Resonance Ionization Mass Spectrometry

DOE PAGES

Savina, Michael R.; Isselhardt, Brett H.; Kucher, Andrew; ...

2017-05-09

Useful yields from resonance ionization mass spectrometry can be extremely high compared to other mass spectrometry techniques, but uranium analysis shows strong matrix effects arising from the tendency of uranium to form strongly bound oxide molecules that do not dissociate appreciably on energetic ion bombardment. Here, we demonstrate a useful yield of 24% for metallic uranium. Modeling the laser ionization and ion transmission processes shows that the high useful yield is attributable to a high ion fraction achieved by resonance ionization. We quantify the reduction of uranium oxide surface layers by Ar + and Ga + sputtering. The useful yieldmore » for uranium atoms from a uranium dioxide matrix is 0.4% and rises to 2% when the surface is in sputter equilibrium with the ion beam. The lower useful yield from the oxide is almost entirely due to uranium oxide molecules reducing the neutral atom content of the sputtered flux. We also demonstrate rapid isotopic analysis of solid uranium oxide at a precision of <0.5% relative standard deviation using relatively broadband lasers to mitigate spectroscopic fractionation.« less

High Useful Yield and Isotopic Analysis of Uranium by Resonance Ionization Mass Spectrometry

DOE Office of Scientific and Technical Information (OSTI.GOV)

Savina, Michael R.; Isselhardt, Brett H.; Kucher, Andrew

Useful yields from resonance ionization mass spectrometry can be extremely high compared to other mass spectrometry techniques, but uranium analysis shows strong matrix effects arising from the tendency of uranium to form strongly bound oxide molecules that do not dissociate appreciably on energetic ion bombardment. Here, we demonstrate a useful yield of 24% for metallic uranium. Modeling the laser ionization and ion transmission processes shows that the high useful yield is attributable to a high ion fraction achieved by resonance ionization. We quantify the reduction of uranium oxide surface layers by Ar + and Ga + sputtering. The useful yieldmore » for uranium atoms from a uranium dioxide matrix is 0.4% and rises to 2% when the surface is in sputter equilibrium with the ion beam. The lower useful yield from the oxide is almost entirely due to uranium oxide molecules reducing the neutral atom content of the sputtered flux. We also demonstrate rapid isotopic analysis of solid uranium oxide at a precision of <0.5% relative standard deviation using relatively broadband lasers to mitigate spectroscopic fractionation.« less

Development and Validation of Capabilities to Measure Thermal Properties of Layered Monolithic U-Mo Alloy Plate-Type Fuel

NASA Astrophysics Data System (ADS)

Burkes, Douglas E.; Casella, Andrew M.; Buck, Edgar C.; Casella, Amanda J.; Edwards, Matthew K.; MacFarlan, Paul J.; Pool, Karl N.; Smith, Frances N.; Steen, Franciska H.

2014-07-01

The uranium-molybdenum (U-Mo) alloy in a monolithic form has been proposed as one fuel design capable of converting some of the world's highest power research reactors from the use of high enriched uranium to low enriched uranium. One aspect of the fuel development and qualification process is to demonstrate appropriate understanding of the thermal-conductivity behavior of the fuel system as a function of temperature and expected irradiation conditions. The purpose of this paper is to verify functionality of equipment installed in hot cells for eventual measurements on irradiated uranium-molybdenum (U-Mo) monolithic fuel specimens, refine procedures to operate the equipment, and validate models to extract the desired thermal properties. The results presented here demonstrate the adequacy of the equipment, procedures, and models that have been developed for this purpose based on measurements conducted on surrogate depleted uranium-molybdenum (DU-Mo) alloy samples containing a Zr diffusion barrier and clad in aluminum alloy 6061 (AA6061). The results are in excellent agreement with thermal property data reported in the literature for similar U-Mo alloys as a function of temperature.

A Point Kinetics Model for Estimating Neutron Multiplication of Bare Uranium Metal in Tagged Neutron Measurements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tweardy, Matthew C.; McConchie, Seth; Hayward, Jason P.

An extension of the point kinetics model is developed in this paper to describe the neutron multiplicity response of a bare uranium object under interrogation by an associated particle imaging deuterium-tritium (D-T) measurement system. This extended model is used to estimate the total neutron multiplication of the uranium. Both MCNPX-PoliMi simulations and data from active interrogation measurements of highly enriched and depleted uranium geometries are used to evaluate the potential of this method and to identify the sources of systematic error. The detection efficiency correction for measured coincidence response is identified as a large source of systematic error. If themore » detection process is not considered, results suggest that the method can estimate total multiplication to within 13% of the simulated value. Values for multiplicity constants in the point kinetics equations are sensitive to enrichment due to (n, xn) interactions by D-T neutrons and can introduce another significant source of systematic bias. This can theoretically be corrected if isotopic composition is known a priori. Finally, the spatial dependence of multiplication is also suspected of introducing further systematic bias for high multiplication uranium objects.« less

A Point Kinetics Model for Estimating Neutron Multiplication of Bare Uranium Metal in Tagged Neutron Measurements

DOE PAGES

Tweardy, Matthew C.; McConchie, Seth; Hayward, Jason P.

2017-06-13

An extension of the point kinetics model is developed in this paper to describe the neutron multiplicity response of a bare uranium object under interrogation by an associated particle imaging deuterium-tritium (D-T) measurement system. This extended model is used to estimate the total neutron multiplication of the uranium. Both MCNPX-PoliMi simulations and data from active interrogation measurements of highly enriched and depleted uranium geometries are used to evaluate the potential of this method and to identify the sources of systematic error. The detection efficiency correction for measured coincidence response is identified as a large source of systematic error. If themore » detection process is not considered, results suggest that the method can estimate total multiplication to within 13% of the simulated value. Values for multiplicity constants in the point kinetics equations are sensitive to enrichment due to (n, xn) interactions by D-T neutrons and can introduce another significant source of systematic bias. This can theoretically be corrected if isotopic composition is known a priori. Finally, the spatial dependence of multiplication is also suspected of introducing further systematic bias for high multiplication uranium objects.« less

HEU Holdup Measurements on 321-M A-Lathe

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dewberry, R.A.

The Analytical Development Section of SRTC was requested by the Facilities Disposition Division (FDD) of the Savannah River Site to determine the holdup of enriched uranium in the 321-M facility as part of an overall deactivation project of the facility. The 321-M facility was used to fabricate enriched uranium fuel assemblies, lithium-aluminum target tubes, neptunium assemblies, and miscellaneous components for the production reactors. The results of the holdup assays are essential for determining compliance with the solid waste Waste Acceptance Criteria, Material Control and Accountability, and to meet criticality safety controls. Three measurement systems were used to determine highly enrichedmore » uranium (HEU) holdup. This report covers holdup measurements on the A-Lathe that was used to machine uranium-aluminum-alloy (U-Al). Our results indicated that the lathe contained more than the limits stated in the Waste Acceptance Criteria (WAC) for the solid waste E-Area Vaults. Thus the lathe was decontaminated three times and assayed four times in order to bring the amounts of uranium to an acceptable content. This report will discuss the methodology, Non-Destructive Assay (NDA) measurements, and results of the U-235 holdup on the lathe.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stillman, J. A.; Feldman, E. E.; Wilson, E. H.

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

10 CFR Appendix H to Part 110 - Illustrative List of Electromagnetic Enrichment Plant Equipment and Components Under NRC Export...

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Illustrative List of Electromagnetic Enrichment Plant... Appendix H to Part 110—Illustrative List of Electromagnetic Enrichment Plant Equipment and Components Under NRC Export Licensing Authority Note—In the electromagnetic process, uranium metal ions produced by...

10 CFR Appendix H to Part 110 - Illustrative List of Electromagnetic Enrichment Plant Equipment and Components Under NRC Export...

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 2 2013-01-01 2013-01-01 false Illustrative List of Electromagnetic Enrichment Plant... Appendix H to Part 110—Illustrative List of Electromagnetic Enrichment Plant Equipment and Components Under NRC Export Licensing Authority Note—In the electromagnetic process, uranium metal ions produced by...

10 CFR Appendix H to Part 110 - Illustrative List of Electromagnetic Enrichment Plant Equipment and Components Under NRC Export...

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Illustrative List of Electromagnetic Enrichment Plant... Appendix H to Part 110—Illustrative List of Electromagnetic Enrichment Plant Equipment and Components Under NRC Export Licensing Authority Note—In the electromagnetic process, uranium metal ions produced by...

10 CFR Appendix H to Part 110 - Illustrative List of Electromagnetic Enrichment Plant Equipment and Components Under NRC Export...

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 2 2012-01-01 2012-01-01 false Illustrative List of Electromagnetic Enrichment Plant... Appendix H to Part 110—Illustrative List of Electromagnetic Enrichment Plant Equipment and Components Under NRC Export Licensing Authority Note—In the electromagnetic process, uranium metal ions produced by...

10 CFR Appendix H to Part 110 - Illustrative List of Electromagnetic Enrichment Plant Equipment and Components Under NRC Export...

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 2 2014-01-01 2014-01-01 false Illustrative List of Electromagnetic Enrichment Plant... Appendix H to Part 110—Illustrative List of Electromagnetic Enrichment Plant Equipment and Components Under NRC Export Licensing Authority Note: In the electromagnetic process, uranium metal ions produced by...

Limiting Regret: Building the Army We Will Need

DTIC Science & Technology

2015-08-18

Recently, U.S. and Chinese experts have estimated that the North Koreans may be able to produce enough fissionable plutonium and uranium to build up...long-range missiles, but their recently revealed ability to separate uranium could give them the ability to build gun-assembled fission weapons similar...weapons programs and living up to their international obligations.” 36North Korea has had a uranium enrichment capacity since at least November 2010

235U Holdup Measurements in the 321-M Exhaust Elbows

DOE Office of Scientific and Technical Information (OSTI.GOV)

Salaymeh, S.R.

The Analytical Development Section of Savannah River Technology Center (SRTC) was requested by the Facilities Disposition Division (FDD) to determine the holdup of enriched uranium in the 321-M facility as part of an overall deactivation project of the facility. The 321-M facility was used to fabricate enriched uranium fuel assemblies, lithium-aluminum target tubes, neptunium assemblies, and miscellaneous components for the production reactors. The facility also includes the 324-M storage building and the passageway connecting it to 321-M. The results of the holdup assays are essential for determining compliance with the Waste Acceptance Criteria, Material Control and Accountability, and to meetmore » criticality safety controls. This report covers holdup measurements of uranium residue in the exhaust piping elbows removed from the roof the 321-M facility.« less

Laser Shockwave Technique For Characterization Of Nuclear Fuel Plate Interfaces

DOE Office of Scientific and Technical Information (OSTI.GOV)

James A. Smith; Barry H. Rabin; Mathieu Perton

2012-07-01

The US National Nuclear Security Agency is tasked with minimizing the worldwide use of high-enriched uranium. One aspect of that effort is the conversion of research reactors to monolithic fuel plates of low-enriched uranium. The manufacturing process includes hot isostatic press bonding of an aluminum cladding to the fuel foil. The Laser Shockwave Technique (LST) is here evaluated for characterizing the interface strength of fuel plates using depleted Uranium/Mo foils. LST is a non-contact method that uses lasers for the generation and detection of large amplitude acoustic waves and is therefore well adapted to the quality assurance of this process.more » Preliminary results show a clear signature of well-bonded and debonded interfaces and the method is able to classify/rank the bond strength of fuel plates prepared under different HIP conditions.« less

Laser shockwave technique for characterization of nuclear fuel plate interfaces

DOE Office of Scientific and Technical Information (OSTI.GOV)

Perton, M.; Levesque, D.; Monchalin, J.-P.

2013-01-25

The US National Nuclear Security Agency is tasked with minimizing the worldwide use of high-enriched uranium. One aspect of that effort is the conversion of research reactors to monolithic fuel plates of low-enriched uranium. The manufacturing process includes hot isostatic press bonding of an aluminum cladding to the fuel foil. The Laser Shockwave Technique (LST) is here evaluated for characterizing the interface strength of fuel plates using depleted Uranium/Mo foils. LST is a non-contact method that uses lasers for the generation and detection of large amplitude acoustic waves and is therefore well adapted to the quality assurance of this process.more » Preliminary results show a clear signature of well-bonded and debonded interfaces and the method is able to classify/rank the bond strength of fuel plates prepared under different HIP conditions.« less

10 CFR 150.11 - Critical mass.

Code of Federal Regulations, 2013 CFR

2013-01-01

... uranium enriched in the isotope U-235 in quantities not exceeding 350 grams of contained U-235; uranium-233 in quantities not exceeding 200 grams; plutonium in quantities not exceeding 200 grams; or any... not exceed the limitation and are within the formula, as follows: (175 (grams contained U-235/350)+(50...

10 CFR 150.11 - Critical mass.

Code of Federal Regulations, 2011 CFR

2011-01-01

... uranium enriched in the isotope U-235 in quantities not exceeding 350 grams of contained U-235; uranium-233 in quantities not exceeding 200 grams; plutonium in quantities not exceeding 200 grams; or any... not exceed the limitation and are within the formula, as follows: (175 (grams contained U-235/350)+(50...

10 CFR 150.11 - Critical mass.

Code of Federal Regulations, 2010 CFR

2010-01-01

... uranium enriched in the isotope U-235 in quantities not exceeding 350 grams of contained U-235; uranium-233 in quantities not exceeding 200 grams; plutonium in quantities not exceeding 200 grams; or any... not exceed the limitation and are within the formula, as follows: (175 (grams contained U-235/350)+(50...

10 CFR 150.11 - Critical mass.

Code of Federal Regulations, 2014 CFR

2014-01-01

... uranium enriched in the isotope U-235 in quantities not exceeding 350 grams of contained U-235; uranium-233 in quantities not exceeding 200 grams; plutonium in quantities not exceeding 200 grams; or any... not exceed the limitation and are within the formula, as follows: (175 (grams contained U-235/350)+(50...

10 CFR 150.11 - Critical mass.

Code of Federal Regulations, 2012 CFR

2012-01-01

... uranium enriched in the isotope U-235 in quantities not exceeding 350 grams of contained U-235; uranium-233 in quantities not exceeding 200 grams; plutonium in quantities not exceeding 200 grams; or any... not exceed the limitation and are within the formula, as follows: (175 (grams contained U-235/350)+(50...

Nonproliferation Challenges in Space Defense Technology - PANEL

NASA Technical Reports Server (NTRS)

Houts, Michael G.

2016-01-01

The use of highly enriched uranium (HEU) almost always "helps" space fission systems. Nuclear Thermal Propulsion (NTP) and high power fission electric systems appear able to use < 20% enriched uranium with minimal / acceptable performance impacts. However, lower power, "entry level" systems may be needed for space fission technology to be developed and utilized. Low power (i.e. approx.1 kWe) fission systems may have an unacceptable performance penalty if LEU is used instead of HEU. Are there Ways to Support Non-Proliferation Objectives While Simultaneously Helping Enable the Development and Utilization of Modern Space Fission Power and Propulsion Systems?

Surplus Highly Enriched Uranium Disposition Program plan

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

1996-10-01

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

KINETICS OF THE DISSOLUTION OF URANIUM DIOXIDE IN CARBONATE-BICARBONATE SOLUTIONS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schortmann, W.E.; DeSesa, M.A.

The kinetics of the dissolution of uranium dioxide in sodium carbonate- sodium bicarbonate solutions were determined. The study was undertaken in order to obtain fundamental information about the commercial carbonate process for leaching uranium from its ores. A rate equation incorporating the effects of surface area oxygen partial pressure, temperature, and reagent concentrations was empirically developed. A mechanism consisting essentially of two consecutive reactions at steady state is proposed. These reactions are the oxidation of U/ sup 4+/ to U/sup 6+/ and the subsequent formation of the uranyl dicarbonate complexion. Depending on the conditions, either or both of these reactionsmore » can determine the over-all rate. The conversion of uranyl dicarbonate to the uranyl tricarbonate complexion is postulated to be very rapid. In the suggested mechanism, the rate-determining phase of the oxidation is the dissociation of adsorbed molecular oxygen. and both the carbonate and bicarbonate ions play equivalent roles in the formation of the uranyl dicarbonate. As indicated by their high activation energies of about 13 and 14 kcal per mole uranium, both reactions are chemical rather than diffusional processes. A mathematical examination of the proposed mechanism produced a rate equation consistent with the experimental information. The credibility of the mechanism was thereby strengthened. (auth)« less

Migration of defect clusters and xenon-vacancy clusters in uranium dioxide

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chen, Dong; Gao, Fei; Deng, Huiqiu

2014-07-01

The possible transition states, minimum energy paths and migration mechanisms of defect clusters and xenon-vacancy defect clusters in uranium dioxide have been investigated using the dimer and the nudged elastic-band methods. The nearby O atom can easily hop into the oxygen vacancy position by overcoming a small energy barrier, which is much lower than that for the migration of a uranium vacancy. A simulation for a vacancy cluster consisting of two oxygen vacancies reveals that the energy barrier of the divacancy migration tends to decrease with increasing the separation distance of divacancy. For an oxygen interstitial, the migration barrier formore » the hopping mechanism is almost three times larger than that for the exchange mechanism. Xe moving between two interstitial sites is unlikely a dominant migration mechanism considering the higher energy barrier. A net migration process of a Xe-vacancy pair containing an oxygen vacancy and a xenon interstitial is identified by the NEB method. We expect the oxygen vacancy-assisted migration mechanism to possibly lead to a long distance migration of the Xe interstitials in UO2. The migration of defect clusters involving Xe substitution indicates that Xe atom migrating away from the uranium vacancy site is difficult.« less

On the nature of the phase transition in uranium dioxide

NASA Astrophysics Data System (ADS)

Gofryk, K.; Mast, D.; Antonio, D.; Shrestha, K.; Andersson, D.; Stanek, C.; Jaime, M.

Uranium dioxide (UO2) is by far the most studied actinide material as it is a primary fuel used in light water nuclear reactors. Its thermal and magnetic properties remain, however, a puzzle resulting from strong couplings between magnetism and lattice vibrations. UO2 crystalizes in the face-centered-cubic fluorite structure and is a Mott-Hubbard insulator with well-localized uranium 5 f-electrons. In addition, below 30 K, a long range antiferromagnetic ordering of the electric-quadrupole of the uranium moments is observed, forming complex non-collinear 3-k magnetic structure. This transition is accompanied by Jahn-Teller distortion of oxygen atoms. It is believed that the first order nature of the transition results from the competition between the exchange interaction and the Jahn-Teller distortion. Here we present results of our extensive thermodynamic investigations on well-characterized and oriented single crystals of UO2+x (x = 0, 0.033, 0.04, and 0.11). By focusing on the transition region under applied magnetic field we are able to study the interplay between different competing interactions (structural, magnetic, and electrical), its dynamics, and relationship to the oxygen content. We will discuss implications of these results. Work supported by the Department of Energy, Office of Basic Energy Sciences, Materials Sciences, and Engineering Division.

21 CFR 137.185 - Enriched self-rising flour.

Code of Federal Regulations, 2011 CFR

2011-04-01

... 21 Food and Drugs 2 2011-04-01 2011-04-01 false Enriched self-rising flour. 137.185 Section 137... Cereal Flours and Related Products § 137.185 Enriched self-rising flour. Enriched self-rising flour... carbon dioxide evolved under ordinary conditions of use of the enriched self-rising flour is not less...

21 CFR 137.185 - Enriched self-rising flour.

Code of Federal Regulations, 2013 CFR

2013-04-01

... 21 Food and Drugs 2 2013-04-01 2013-04-01 false Enriched self-rising flour. 137.185 Section 137... Cereal Flours and Related Products § 137.185 Enriched self-rising flour. Enriched self-rising flour... carbon dioxide evolved under ordinary conditions of use of the enriched self-rising flour is not less...

21 CFR 137.185 - Enriched self-rising flour.

Code of Federal Regulations, 2012 CFR

2012-04-01

... 21 Food and Drugs 2 2012-04-01 2012-04-01 false Enriched self-rising flour. 137.185 Section 137... Cereal Flours and Related Products § 137.185 Enriched self-rising flour. Enriched self-rising flour... carbon dioxide evolved under ordinary conditions of use of the enriched self-rising flour is not less...

21 CFR 137.185 - Enriched self-rising flour.

Code of Federal Regulations, 2014 CFR

2014-04-01

... 21 Food and Drugs 2 2014-04-01 2014-04-01 false Enriched self-rising flour. 137.185 Section 137... Cereal Flours and Related Products § 137.185 Enriched self-rising flour. Enriched self-rising flour... carbon dioxide evolved under ordinary conditions of use of the enriched self-rising flour is not less...

Illicit Trafficking of Natural Radionuclides

DOE Office of Scientific and Technical Information (OSTI.GOV)

Friedrich, Steinhaeusler; Lyudmila, Zaitseva

2008-08-07

Natural radionuclides have been subject to trafficking worldwide, involving natural uranium ore (U 238), processed uranium (yellow cake), low enriched uranium (<20% U 235) or highly enriched uranium (>20% U 235), radium (Ra 226), polonium (Po 210), and natural thorium ore (Th 232). An important prerequisite to successful illicit trafficking activities is access to a suitable logistical infrastructure enabling an undercover shipment of radioactive materials and, in case of trafficking natural uranium or thorium ore, capable of transporting large volumes of material. Covert en route diversion of an authorised uranium transport, together with covert diversion of uranium concentrate from anmore » operating or closed uranium mines or mills, are subject of case studies. Such cases, involving Israel, Iran, Pakistan and Libya, have been analyzed in terms of international actors involved and methods deployed. Using international incident data contained in the Database on Nuclear Smuggling, Theft and Orphan Radiation Sources (DSTO) and international experience gained from the fight against drug trafficking, a generic Trafficking Pathway Model (TPM) is developed for trafficking of natural radionuclides. The TPM covers the complete trafficking cycle, ranging from material diversion, covert material transport, material concealment, and all associated operational procedures. The model subdivides the trafficking cycle into five phases: (1) Material diversion by insider(s) or initiation by outsider(s); (2) Covert transport; (3) Material brokerage; (4) Material sale; (5) Material delivery. An Action Plan is recommended, addressing the strengthening of the national infrastructure for material protection and accounting, development of higher standards of good governance, and needs for improving the control system deployed by customs, border guards and security forces.« less

Illicit Trafficking of Natural Radionuclides

NASA Astrophysics Data System (ADS)

Friedrich, Steinhäusler; Lyudmila, Zaitseva

2008-08-01

Natural radionuclides have been subject to trafficking worldwide, involving natural uranium ore (U 238), processed uranium (yellow cake), low enriched uranium (<20% U 235) or highly enriched uranium (>20% U 235), radium (Ra 226), polonium (Po 210), and natural thorium ore (Th 232). An important prerequisite to successful illicit trafficking activities is access to a suitable logistical infrastructure enabling an undercover shipment of radioactive materials and, in case of trafficking natural uranium or thorium ore, capable of transporting large volumes of material. Covert en route diversion of an authorised uranium transport, together with covert diversion of uranium concentrate from an operating or closed uranium mines or mills, are subject of case studies. Such cases, involving Israel, Iran, Pakistan and Libya, have been analyzed in terms of international actors involved and methods deployed. Using international incident data contained in the Database on Nuclear Smuggling, Theft and Orphan Radiation Sources (DSTO) and international experience gained from the fight against drug trafficking, a generic Trafficking Pathway Model (TPM) is developed for trafficking of natural radionuclides. The TPM covers the complete trafficking cycle, ranging from material diversion, covert material transport, material concealment, and all associated operational procedures. The model subdivides the trafficking cycle into five phases: (1) Material diversion by insider(s) or initiation by outsider(s); (2) Covert transport; (3) Material brokerage; (4) Material sale; (5) Material delivery. An Action Plan is recommended, addressing the strengthening of the national infrastructure for material protection and accounting, development of higher standards of good governance, and needs for improving the control system deployed by customs, border guards and security forces.

The separation of lanthanides and actinides in supercritical fluid carbon dioxide

DOE PAGES

Mincher, Bruce J.; Wai, Chien M.; Fox, Robert V.; ...

2015-10-28

Supercritical fluid carbon dioxide presents an attractive alternative to conventional solvents for recovery of the actinides and lanthanides. Carbon dioxide is a good solvent for fluorine and phosphate-containing ligands, including the traditional tributylphosphate ligand used in process-scale uranium separations. Actinide and lanthanide oxides may even be directly dissolved in carbon dioxide containing the complexes formed between these ligands and mineral acids, obviating the need for large volumes of acids for leaching and dissolution, and the corresponding organic liquid–liquid solvent extraction solutions. As a result, examples of the application of this novel technology for actinide and lanthanide separations are presented.

Photocatalytic decomposition of Rhodamine B on uranium-doped mesoporous titanium dioxide

DOE PAGES

Liu, Yi; Becker, Blake; Burdine, Brandon; ...

2017-04-13

Mesoporous uranium-doped TiO 2 anatase materials were studied to determine the influence of U-doping on the photocatalytic properties for Rhodamine B (RhB) degradation. The physico-chemical properties of the samples were characterized and the results of X-ray diffraction, transmission electron microscopy, and Raman spectroscopy demonstrate homogeneous incorporation of uranium into the anatase lattice. X-ray photoelectron spectroscopy of the doped anatase confirmed the dominance of the U 4+ species and an increasing proportion of U 6+ species as the uranium doping was increased. The absorption thresholds of the uranium-doped anatase extended into the visible light region. A synergistic effect of the bandmore » gap energy and oxidation state of the dopant contribute to an enhanced photocatalytic capability for RhB degradation by U-doped TiO 2.« less

Photocatalytic decomposition of Rhodamine B on uranium-doped mesoporous titanium dioxide

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liu, Yi; Becker, Blake; Burdine, Brandon

Mesoporous uranium-doped TiO 2 anatase materials were studied to determine the influence of U-doping on the photocatalytic properties for Rhodamine B (RhB) degradation. The physico-chemical properties of the samples were characterized and the results of X-ray diffraction, transmission electron microscopy, and Raman spectroscopy demonstrate homogeneous incorporation of uranium into the anatase lattice. X-ray photoelectron spectroscopy of the doped anatase confirmed the dominance of the U 4+ species and an increasing proportion of U 6+ species as the uranium doping was increased. The absorption thresholds of the uranium-doped anatase extended into the visible light region. A synergistic effect of the bandmore » gap energy and oxidation state of the dopant contribute to an enhanced photocatalytic capability for RhB degradation by U-doped TiO 2.« less

Proteome changes in rat serum after a chronic ingestion of enriched uranium: Toward a biological signature of internal contamination and radiological effect.

PubMed

Petitot, F; Frelon, S; Chambon, C; Paquet, F; Guipaud, O

2016-08-22

The civilian and military use of uranium results in an increased risk of human exposure. The toxicity of uranium results from both its chemical and radiological properties that vary with isotopic composition. Validated biomarkers of health effects associated with exposure to uranium are neither sensitive nor specific to uranium radiotoxicity and/or radiological effect. This study aimed at investigating if serum proteins could be useful as biomarkers of both uranium exposure and radiological effect. Male Sprague-Dawley rats were chronically exposed through drinking water to low levels (40mg/L, corresponding to 1mg of uranium per animal per day) of either 4% (235)U-enriched uranium (EU) or 12% EU during 6 weeks. A proteomics approach based on two-dimensional electrophoresis (2D-DIGE) and mass spectrometry (MS) was used to establish protein expression profiles that could be relevant for discriminating between groups, and to identify some differentially expressed proteins following uranium ingestion. It demonstrated that the expressions of 174 protein spots over 1045 quantified spots were altered after uranium exposure (p<0.05). Using both inferential and non-supervised multivariate statistics, we show sets of spots features that lead to a clear discrimination between controls and EU exposed groups on the one hand (21 spots), and between 4% EU and 12% EU on the other hand (7 spots), showing that investigation of the serum proteome may possibly be of relevance to address both uranium contamination and radiological effect. Finally, using bioinformatics tools, pathway analyses of differentially expressed MS-identified proteins find that acute phase, inflammatory and immune responses as well as oxidative stress are likely involved in the response to contamination, suggesting a physiological perturbation, but that does not necessarily lead to a toxic effect. Copyright © 2016 Elsevier Ireland Ltd. All rights reserved.

Americium characterization by X-ray fluorescence and absorption spectroscopy in plutonium uranium mixed oxide

NASA Astrophysics Data System (ADS)

Degueldre, Claude; Cozzo, Cedric; Martin, Matthias; Grolimund, Daniel; Mieszczynski, Cyprian

2013-06-01

Plutonium uranium mixed oxide (MOX) fuels are currently used in nuclear reactors. The actinides in these fuels need to be analyzed after irradiation for assessing their behaviour with regard to their environment and the coolant. In this work the study of the atomic structure and next-neighbour environment of Am in the (Pu,U)O2 lattice in an irradiated (60 MW d kg-1) MOX sample was performed employing micro-X-ray fluorescence (µ-XRF) and micro-X-ray absorption fine structure (µ-XAFS) spectroscopy. The chemical bonds, valences and stoichiometry of Am (˜0.66 wt%) are determined from the experimental data gained for the irradiated fuel material examined in its peripheral zone (rim) of the fuel. In the irradiated sample Am builds up as Am3+ species within an [AmO8]13- coordination environment (e.g. >90%) and no (<10%) Am(IV) or (V) can be detected in the rim zone. The occurrence of americium dioxide is avoided by the redox buffering activity of the uranium dioxide matrix.

10 CFR 766.103 - Special Assessment invoices.

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 4 2010-01-01 2010-01-01 false Special Assessment invoices. 766.103 Section 766.103 Energy DEPARTMENT OF ENERGY URANIUM ENRICHMENT DECONTAMINATION AND DECOMMISSIONING FUND; PROCEDURES FOR... will specify itemized quantities of enrichment services by reactor. In each Special Assessment invoice...

SEPARATION OF URANIUM FROM THORIUM AND PROTACTINIUM

DOEpatents

Musgrave, W.K.R.

1959-06-30

This patent relates to the separation of uranium from thorium and protactinium; such mixtures of elements usually being obtained by neutron irradiation of thorium. The method of separating the constituents has been first to dissolve the mixture of elements in concertrated nitric acid and then to remove the protactinium by absorption on manganese dioxide and the uranium by solvent extraction with ether. Prior to now, comparatively large amounts of thorium were extracted with the uranium. According to the invention this is completely prevented by adding sodium diethyldithiocarbamate to the mixture of soluble nitrate salts. The organic salt has the effect of reacting only with the uranyl nitrate to form the corresponding uranyl salt which can then be selectively extracted from the mixture with amyl acetate.

10 CFR Appendix G to Part 110 - Illustrative List of Plasma Separation Enrichment Plant Equipment and Components Under NRC Export...

Code of Federal Regulations, 2010 CFR

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Illustrative List of Plasma Separation Enrichment Plant... Appendix G to Part 110—Illustrative List of Plasma Separation Enrichment Plant Equipment and Components Under NRC Export Licensing Authority Note—In the plasma separation process, a plasma of uranium ions...

10 CFR Appendix G to Part 110 - Illustrative List of Plasma Separation Enrichment Plant Equipment and Components Under NRC Export...

Code of Federal Regulations, 2012 CFR

2012-01-01

... 10 Energy 2 2012-01-01 2012-01-01 false Illustrative List of Plasma Separation Enrichment Plant... Appendix G to Part 110—Illustrative List of Plasma Separation Enrichment Plant Equipment and Components Under NRC Export Licensing Authority Note—In the plasma separation process, a plasma of uranium ions...

10 CFR Appendix G to Part 110 - Illustrative List of Plasma Separation Enrichment Plant Equipment and Components Under NRC Export...

Code of Federal Regulations, 2011 CFR

2011-01-01

... 10 Energy 2 2011-01-01 2011-01-01 false Illustrative List of Plasma Separation Enrichment Plant... Appendix G to Part 110—Illustrative List of Plasma Separation Enrichment Plant Equipment and Components Under NRC Export Licensing Authority Note—In the plasma separation process, a plasma of uranium ions...

10 CFR Appendix G to Part 110 - Illustrative List of Plasma Separation Enrichment Plant Equipment and Components Under NRC Export...

Code of Federal Regulations, 2013 CFR

2013-01-01

... 10 Energy 2 2013-01-01 2013-01-01 false Illustrative List of Plasma Separation Enrichment Plant... Appendix G to Part 110—Illustrative List of Plasma Separation Enrichment Plant Equipment and Components Under NRC Export Licensing Authority Note—In the plasma separation process, a plasma of uranium ions...

10 CFR Appendix G to Part 110 - Illustrative List of Plasma Separation Enrichment Plant Equipment and Components Under NRC Export...

Code of Federal Regulations, 2014 CFR

2014-01-01

... 10 Energy 2 2014-01-01 2014-01-01 false Illustrative List of Plasma Separation Enrichment Plant... Appendix G to Part 110—Illustrative List of Plasma Separation Enrichment Plant Equipment and Components Under NRC Export Licensing Authority Note: In the plasma separation process, a plasma of uranium ions...

Recovery of uranium from seawater by immobilized tannin

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sakaguchi, T.; Nakajima, A.

1987-06-01

Tannin compounds having multiple adjacent hydroxy groups have an extremely high affinity for uranium. To prevent the leaching of tannins into water and to improve the adsorbing characteristics of these compounds, the authors tried to immobilize tannins. The immobilized tannin has the most favorable features for uranium recovery; high selective adsorption ability to uranium, rapid adsorption rate, and applicability in both column and batch systems. The immobilized tannin can recover uranium from natural seawater with high efficiency. About 2530 ..mu..g uranium is adsorbed per gram of this adsorbent within 22 h. Depending on the concentration in seawater, an enrichment ofmore » up to 766,000-fold within the adsorbent is possible. Almost all uranium adsorbed is easily desorbed with a very dilute acid. Thus, the immobilized tannin can be used repeatedly in the adsorption-desorption process.« less

PROCESSES OF RECLAIMING URANIUM FROM SOLUTIONS

DOEpatents

Zumwalt, L.R.

1959-02-10

A process is described for reclaiming residual enriched uranium from calutron wash solutions containing Fe, Cr, Cu, Ni, and Mn as impurities. The solution is adjusted to a pH of between 2 and 4 and is contacted with a metallic reducing agent, such as iron or zinc, in order to reduce the copper to metal and thereby remove it from the solution. At the same time the uranium present is reduced to the uranous state The solution is then contacted with a precipitate of zinc hydroxide or barium carbonate in order to precipitate and carry uranium, iron, and chromium away from the nickel and manganese ions in the solution. The uranium is then recovered fronm this precipitate.

U.S.-Australia Civilian Nuclear Cooperation: Issues for Congress

DTIC Science & Technology

2010-07-07

Mining and Milling ................................................................................................7 Uranium Sales to India...carried out at Lucas Heights (see below). The nuclear fuel cycle begins with mining uranium ore and upgrading it to yellowcake. Because naturally... mining and milling stage. Commercial enrichment services are available in the United States, Europe, Russia, and Japan. Fuel fabrication services are

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

Code of Federal Regulations, 2010 CFR

2010-01-01

... and special nuclear material in the accounting records are based on measured values; (3) A measurement... 10 Energy 2 2010-01-01 2010-01-01 false Nuclear material control and accounting for uranium... Section 74.33 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) MATERIAL CONTROL AND ACCOUNTING OF SPECIAL...

Analysis of irradiated U-7wt%Mo dispersion fuel microstructures using automated image processing

DOE PAGES

Collette, R.; King, J.; Buesch, C.; ...

2016-04-01

The High Performance Research Reactor Fuel Development (HPPRFD) program is responsible for developing low enriched uranium (LEU) fuel substitutes for high performance reactors fueled with highly enriched uranium (HEU) that have not yet been converted to LEU. The uranium-molybdenum (U-Mo) fuel system was selected for this effort. In this study, fission gas pore segmentation was performed on U-7wt%Mo dispersion fuel samples at three separate fission densities using an automated image processing interface developed in MATLAB. Pore size distributions were attained that showed both expected and unexpected fission gas behavior. In general, it proved challenging to identify any dominant trends whenmore » comparing fission bubble data across samples from different fuel plates due to varying compositions and fabrication techniques. Here, the results exhibited fair agreement with the fission density vs. porosity correlation developed by the Russian reactor conversion program.« less

Analysis of irradiated U-7wt%Mo dispersion fuel microstructures using automated image processing

DOE Office of Scientific and Technical Information (OSTI.GOV)

Collette, R.; King, J.; Buesch, C.

The High Performance Research Reactor Fuel Development (HPPRFD) program is responsible for developing low enriched uranium (LEU) fuel substitutes for high performance reactors fueled with highly enriched uranium (HEU) that have not yet been converted to LEU. The uranium-molybdenum (U-Mo) fuel system was selected for this effort. In this study, fission gas pore segmentation was performed on U-7wt%Mo dispersion fuel samples at three separate fission densities using an automated image processing interface developed in MATLAB. Pore size distributions were attained that showed both expected and unexpected fission gas behavior. In general, it proved challenging to identify any dominant trends whenmore » comparing fission bubble data across samples from different fuel plates due to varying compositions and fabrication techniques. Here, the results exhibited fair agreement with the fission density vs. porosity correlation developed by the Russian reactor conversion program.« less

Spectral pathways for exploration of secondary uranium: An investigation in the desertic tracts of Rajasthan and Gujarat, India

NASA Astrophysics Data System (ADS)

Bharti, Rishikesh; Kalimuthu, R.; Ramakrishnan, D.

2015-10-01

This study aims at identifying potential zones of secondary uranium enrichment using hyperspectral remote sensing, γ-ray spectrometry, fluorimetry and geochemical techniques in the western Rajasthan and northern Gujarat, India. The investigated area has suitable source rocks, conducive past-, and present-climate that can facilitate such enrichment. This enrichment process involves extensive weathering of uranium bearing source rocks, leaching of uranyl compounds in groundwater, and their precipitation in chemical deltas along with duricrusts like calcretes and gypcretes. Spatial distribution of groundwater calcretes (that are rich in Mg-calcite) and gypcretes (that are rich in gypsum) along palaeochannels and chemical deltas were mapped using hyperspectral remote sensing data based on spectral absorptions in 1.70 μm, 2.16 μm, 2.21 μm, 2.33 μm, 2.44 μm wavelength regions. Subsequently based on field radiometric survey, zones of U anomalies were identified and samples of duricrusts and groundwater were collected for geochemical analyses. Anomalous concentration of U (2345.7 Bq/kg) and Th (142.3 Bq/kg) are observed in both duricrusts and groundwater (U-1791 μg/l, Th-34 μg/l) within the palaeo-delta and river confluence. The estimated carnotite Solubility Index also indicates the secondary enrichment of U and the likelihood of occurrence of an unconventional deposit.

Analysis of Tank 38H (HTF-38-15-119, 127) Surface, Subsurface and Tank 43H (HTF-43-15-116, 117 and 118) Surface, Feed Pump Suction and Jet Suction Subsurface Supernatant Samples in Support of Enrichment, Corrosion Control and Salt Batch Planning Programs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Oji, L.

Compositional feed limits have been established to ensure that a nuclear criticality event for the 2H and 3H Evaporators is not possible. The Enrichment Control Program (ECP) requires feed sampling to determine the equivalent enriched uranium content prior to transfer of waste other than recycle transfers (requires sampling to determine the equivalent enriched uranium at two locations in Tanks 38H and 43H every 26 weeks) The Corrosion Control Program (CCP) establishes concentration and temperature limits for key constituents and periodic sampling and analysis to confirm that waste supernate is within these limits. This report provides the results of analyses onmore » Tanks 38H and 43H surface and subsurface supernatant liquid samples in support of the ECP, the CCP, and the Salt Batch 10 Planning Program.« less

Measures of the environmental footprint of the front end of the nuclear fuel cycle

DOE Office of Scientific and Technical Information (OSTI.GOV)

E. Schneider; B. Carlsen; E. Tavrides

2013-11-01

Previous estimates of environmental impacts associated with the front end of the nuclear fuel cycle (FEFC) have focused primarily on energy consumption and CO2 emissions. Results have varied widely. This work builds upon reports from operating facilities and other primary data sources to build a database of front end environmental impacts. This work also addresses land transformation and water withdrawals associated with the processes of the FEFC. These processes include uranium extraction, conversion, enrichment, fuel fabrication, depleted uranium disposition, and transportation. To allow summing the impacts across processes, all impacts were normalized per tonne of natural uranium mined as wellmore » as per MWh(e) of electricity produced, a more conventional unit for measuring environmental impacts that facilitates comparison with other studies. This conversion was based on mass balances and process efficiencies associated with the current once-through LWR fuel cycle. Total energy input is calculated at 8.7 x 10- 3 GJ(e)/MWh(e) of electricity and 5.9 x 10- 3 GJ(t)/MWh(e) of thermal energy. It is dominated by the energy required for uranium extraction, conversion to fluoride compound for subsequent enrichment, and enrichment. An estimate of the carbon footprint is made from the direct energy consumption at 1.7 kg CO2/MWh(e). Water use is likewise dominated by requirements of uranium extraction, totaling 154 L/MWh(e). Land use is calculated at 8 x 10- 3 m2/MWh(e), over 90% of which is due to uranium extraction. Quantified impacts are limited to those resulting from activities performed within the FEFC process facilities (i.e. within the plant gates). Energy embodied in material inputs such as process chemicals and fuel cladding is identified but not explicitly quantified in this study. Inclusion of indirect energy associated with embodied energy as well as construction and decommissioning of facilities could increase the FEFC energy intensity estimate by a factor of up to 2.« less

235U Holdup Measurements in Three 321-M Exhaust HEPA Banks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dewberry, R

2005-02-24

The Analytical Development Section of Savannah River National Laboratory (SRNL) was requested by the Facilities Disposition Division to determine the holdup of enriched uranium in the 321-M facility as part of an overall deactivation project of the facility. The 321-M facility was used to fabricate enriched uranium fuel assemblies, lithium-aluminum target tubes, neptunium assemblies, and miscellaneous components for the production reactors. The results of the holdup assays are essential for determining compliance with the Waste Acceptance Criteria, Material Control & Accountability, and to meet criticality safety controls. This report covers holdup measurements of uranium residue in three HEPA filter exhaustmore » banks of the 321-M facility. Each of the exhaust banks has dimensions near 7' x 14' x 4' and represents a complex holdup problem. A portable HPGe detector and EG&G Dart system that contains the high voltage power supply and signal processing electronics were used to determine highly enriched uranium (HEU) holdup. A personal computer with Gamma-Vision software was used to control the Dart MCA and to provide space to store and manipulate multiple 4096-channel {gamma}-ray spectra. Some acquisitions were performed with the portable detector configured to a Canberra Inspector using NDA2000 acquisition and analysis software. Our results for each component uses a mixture of redundant point source and area source acquisitions that yielded HEU contents in the range of 2-10 grams. This report discusses the methodology, non-destructive assay (NDA) measurements, assumptions, and results of the uranium holdup in these items. This report includes use of transmission-corrected assay as well as correction for contributions from secondary area sources.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

McKenzie, IV, George Espy; Goda, Joetta Marie; Grove, Travis Justin

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

Dose and Dose Risk Caused by Natural Phenomena - Proposed Powder Metallurgy Core Manufacturing Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Holmes, W.G.

2001-08-16

The offsite radiological effects from high velocity straight winds, tornadoes, and earthquakes have been estimated for a proposed facility for manufacturing enriched uranium fuel cores by powder metallurgy. Projected doses range up to 30 mrem/event to the maximum offsite individual for high winds and up to 85 mrem/event for very severe earthquakes. Even under conservative assumptions on meteorological conditions, the maximum offsite dose would be about 20 per cent of the DOE limit for accidents involving enriched uranium storage facilities. The total dose risk is low and is dominated by the risk from earthquakes. This report discusses this test.

Effects of long-term (10 years) exposure to elevated CO2 and O3 on trembling Aspen carbon and nitrogen metabolism at the aspen FACE (Free-Air Carbon Dioxide Enrichment) study site [Abstract

Treesearch

R. Minocha; S. Long; S. Minocha; P Marquardt; M. Kubiske

2010-01-01

The objective of the present study was to evaluate the long-term (10 years) effects of elevated CO2 and O3 on the carbon and nitrogen metabolism of aspen trees. The study was conducted at the Aspen Free-Air Carbon Dioxide Enrichment (FACE) experimental site, Rhinelander, WI, (USA).

Effects of long-term (10 years) exposure to elevated CO2 and O3 on trembling Aspen carbon and nitrogen metabolism at the aspen FACE (Free-Air Carbon Dioxide Enrichment) study site

Treesearch

Rakesh Minocha; Stephanie Long; Subhash Minocha; Paula Marquardt; Neil Nelson; Mark. Kubiske

2010-01-01

This study was conducted at the Aspen Free-Air Carbon Dioxide Enrichment (FACE) experimental site, Rhinelander, WI, (USA). Since 1998, 12 experimental rings planted in 1997 underwent four different treatments: control; elevated CO2 (560 ppm); elevated O3 (1.5X ambient) and elevated CO2 (560 ppm) + O...

Recycled Uranium Mass Balance Project Y-12 National Security Complex Site Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

2000-12-01

This report has been prepared to summarize the findings of the Y-12 National Security Complex (Y-12 Complex) Mass Balance Project and to support preparation of associated U. S. Department of Energy (DOE) site reports. The project was conducted in support of DOE efforts to assess the potential for health and environmental issues resulting from the presence of transuranic (TRU) elements and fission products in recycled uranium (RU) processed by DOE and its predecessor agencies. The United States government used uranium in fission reactors to produce plutonium and tritium for nuclear weapons production. Because uranium was considered scarce relative to demandmore » when these operations began almost 50 years ago, the spent fuel from U.S. fission reactors was processed to recover uranium for recycling. The estimated mass balance for highly enriched RU, which is of most concern for worker exposure and is the primary focus of this project, is summarized in a table. A discrepancy in the mass balance between receipts and shipments (plus inventory and waste) reflects an inability to precisely distinguish between RU and non-RU shipments and receipts involving the Y-12 Complex and Savannah River. Shipments of fresh fuel (non-RU) and sweetener (also non-RU) were made from the Y-12 Complex to Savannah River along with RU shipments. The only way to distinguish between these RU and non-RU streams using available records is by enrichment level. Shipments of {le}90% enrichment were assumed to be RU. Shipments of >90% enrichment were assumed to be non-RU fresh fuel or sweetener. This methodology using enrichment level to distinguish between RU and non-RU results in good estimates of RU flows that are reasonably consistent with Savannah River estimates. Although this is the best available means of distinguishing RU streams, this method does leave a difference of approximately 17.3 MTU between receipts and shipments. Slightly depleted RU streams received by the Y-12 Complex from ORGDP and PGDP are believed to have been returned to the shipping site or disposed of as waste on the Oak Ridge Reservation. No evidence of Y-12 Complex processing of this material was identified in the historical records reviewed by the Project Team.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Campbell, Keri; Judge, Elizabeth J.; Dirmyer, Matthew R.

Surrogate nuclear explosive debris was synthesized and characterized for major, minor, and trace elemental composition as well as uranium isotopics. The samples consisted of an urban glass matrix, equal masses soda lime and cement, doped with 500 ppm uranium with varying enrichments. The surface and cross section morphology were measured with SEM, and the major elemental composition was determined by XPS. LA-ICP-MS was used to measure the uranium isotopic abundance comparing different sampling techniques. Furthermore, the results provide an example of the utility of LA-ICP-MS for forensics applications.

Preparation and benchmarking of ANSL-V cross sections for advanced neutron source reactor studies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Arwood, J.W.; Ford, W.E. III; Greene, N.M.

1987-01-01

Validity of selected data from the fine-group neutron library was satisfactorily tested in performance parameter calculations for the BAPL-1, TRX-1, and ZEEP-1 thermal lattice benchmarks. BAPL-2 is an H/sub 2/O moderated, uranium oxide lattice; TRX-1 is an H/sub 2/O moderated, 1.31 weight percent enriched uranium metal lattice; ZEEP-1 is a D/sub 2/O-moderated, natural uranium lattice. 26 refs., 1 tab.

Background and Source Term Identification in Active Neutron Interrogation Methods

DTIC Science & Technology

2011-03-24

interactions occurred to observe gamma ray peaks and not unduly increase simulation time. Not knowing the uranium enrichment modeled by Gozani, pure U...neutron interactions can occur. The uranium targets, though, should have increased neutron fluencies as the energy levels become below 2 MeV. This is...Assessment Monitor Site (TEAMS) at Kirtland AFB, NM. Iron (Fe-56), lead (Pb-207), polyethylene (C2H4 –– > C-12 & H-1), and uranium (U-235 and U-238) were

Russia ties HEU sale to suspension agreement

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1993-11-01

Unless the US government allows the Russians access to the US uranium fuel market, the successful completion of a high-enriched uranium (HEU) sales agreement between the two governments may be in jeopardy. It had been rumored that the Russians, who have been unhappy about the stiff tariffs imposed on former Soviet uranium in the US market, might use the ongoing HEU negotiations with the White House to ease the antidumping tariffs imposed by the Department of Commerce's International Trade Commission.

Uranium: Prices, rise, then fall

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pool, T.C.

Uranium prices hit eight-year highs in both market tiers,more » $$16.60/lb U{sub 3}O{sub 8} for non-former Soviet Union (FSU) origin and $$15.50 for FSU origin during mid 1996. However, they declined to $14.70 and $13.90, respectively, by the end of the year. Increased uranium prices continue to encourage new production and restarts of production facilities presently on standby. Australia scrapped its {open_quotes}three-mine{close_quotes} policy following the ouster of the Labor party in a March election. The move opens the way for increasing competition with Canada`s low-cost producers. Other events in the industry during 1996 that have current or potential impacts on the market include: approval of legislation outlining the ground rules for privatization of the US Enrichment Corp. (USEC) and the subsequent sales of converted Russian highly enriched uranium (HEU) from its nuclear weapons program, announcement of sales plans for converted US HEU and other surplus material through either the Department of Energy or USEC, and continuation of quotas for uranium from the FSU in the United States and Europe. In Canada, permitting activities continued on the Cigar Lake and McArthur River projects; and construction commenced on the McClean Lake mill.« less

Experiments in anodic film effects during electrorefining of scrap U-10Mo fuels in support of modeling efforts

DOE Office of Scientific and Technical Information (OSTI.GOV)

Van Kleeck, M.; Chemical Sciences and Engineering Division, Argonne National Laboratory, Argonne, IL 60439; Willit, J.

A monolithic uranium molybdenum alloy clad in zirconium has been proposed as a low enriched uranium (LEU) fuel option for research and test reactors, as part of the Reduced Enrichment for Research and Test Reactors program. Scrap from the fuel's manufacture will contain a significant portion of recoverable LEU. Pyroprocessing has been identified as an option to perform this recovery. A model of a pyroprocessing recovery procedure has been developed to assist in refining the LEU recovery process and designing the facility. Corrosion theory and a two mechanism transport model were implemented on a Mat-Lab platform to perform the modeling.more » In developing this model, improved anodic behavior prediction became necessary since a dense uranium-rich salt film was observed at the anode surface during electrorefining experiments. Experiments were conducted on uranium metal to determine the film's character and the conditions under which it forms. The electro-refiner salt used in all the experiments was eutectic LiCl/KCl containing UCl{sub 3}. The anodic film material was analyzed with ICP-OES to determine its composition. Both cyclic voltammetry and potentiodynamic scans were conducted at operating temperatures between 475 and 575 C. degrees to interrogate the electrochemical behavior of the uranium. The results show that an anodic film was produced on the uranium electrode. The film initially passivated the surface of the uranium on the working electrode. At high over potentials after a trans-passive region, the current observed was nearly equal to the current observed at the initial active level. Analytical results support the presence of K{sub 2}UCl{sub 6} at the uranium surface, within the error of the analytical method.« less

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

DOEpatents

Travelli, A.

1985-10-25

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

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

DOEpatents

Travelli, Armando

1988-01-01

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

Molten uranium dioxide structure and dynamics

DOE PAGES

Skinner, L. B.; Parise, J. B.; Benmore, C. J.; ...

2014-11-21

Uranium dioxide (UO 2) is the major nuclear fuel component of fission power reactors. A key concern during severe accidents is the melting and leakage of radioactive UO 2 as it corrodes through its zirconium cladding and steel containment. Yet, the very high temperatures (>3140 kelvin) and chemical reactivity of molten UO 2 have prevented structural studies. In this work, we combine laser heating, sample levitation, and synchrotron x-rays to obtain pair distribution function measurements of hot solid and molten UO 2. The hot solid shows a substantial increase in oxygen disorder around the lambda transition (2670 K) but negligiblemore » U-O coordination change. On melting, the average U-O coordination drops from 8 to 6.7 ± 0.5. Molecular dynamics models refined to this structure predict higher U-U mobility than 8-coordinated melts.« less

In-situ, time resolved monitoring of uranium in BFS:OPC grout. Part 1: Corrosion in water vapour.

PubMed

Stitt, C A; Paraskevoulakos, C; Banos, A; Harker, N J; Hallam, K R; Davenport, A; Street, S; Scott, T B

2017-08-11

Uranium encapsulated in grout was exposed to water vapour for extended periods of time. Through synchrotron x-ray powder diffraction and tomography measurements, uranium dioxide was determined the dominant corrosion product over a 50-week time period. The oxide growth rate initiated rapidly, with rates comparable to the U + H 2 O reaction. Over time, the reaction rate decreased and eventually plateaued to a rate similar to the U + H 2 O + O 2 reaction. This behaviour was not attributed to oxygen ingress, but instead the decreasing permeability of the grout, limiting oxidising species access to the metal surface.

Polarized-neutron-scattering study of the spin-wave excitations in the 3-k ordered phase of uranium antimonide.

PubMed

Magnani, N; Caciuffo, R; Lander, G H; Hiess, A; Regnault, L-P

2010-03-24

The anisotropy of magnetic fluctuations propagating along the [1 1 0] direction in the ordered phase of uranium antimonide has been studied using polarized inelastic neutron scattering. The observed polarization behavior of the spin waves is a natural consequence of the longitudinal 3-k magnetic structure; together with recent results on the 3-k-transverse uranium dioxide, these findings establish this technique as an important tool to study complex magnetic arrangements. Selected details of the magnon excitation spectra of USb have also been reinvestigated, indicating the need to revise the currently accepted theoretical picture for this material.

Status Report on the Passive Neutron Enrichment Meter (PNEM) for UF6 Cylinder Assay

DOE Office of Scientific and Technical Information (OSTI.GOV)

Miller, Karen A.; Swinhoe, Martyn T.; Menlove, Howard O.

2012-05-02

The Passive Neutron Enrichment Meter (PNEM) is a nondestructive assay (NDA) system being developed at Los Alamos National Laboratory (LANL). It was designed to determine {sup 235}U mass and enrichment of uranium hexafluoride (UF{sub 6}) in product, feed, and tails cylinders (i.e., 30B and 48Y cylinders). These cylinders are found in the nuclear fuel cycle at uranium conversion, enrichment, and fuel fabrication facilities. The PNEM is a {sup 3}He-based neutron detection system that consists of two briefcase-sized detector pods. A photograph of the system during characterization at LANL is shown in Fig. 1. Several signatures are currently being studied tomore » determine the most effective measurement and data reduction technique for unfolding {sup 235}U mass and enrichment. The system collects total neutron and coincidence data for both bare and cadmium-covered detector pods. The measurement concept grew out of the success of the Uranium Cylinder Assay System (UCAS), which is an operator system at Rokkasho Enrichment Plant (REP) that uses total neutron counting to determine {sup 235}U mass in UF{sub 6} cylinders. The PNEM system was designed with higher efficiency than the UCAS in order to add coincidence counting functionality for the enrichment determination. A photograph of the UCAS with a 48Y cylinder at REP is shown in Fig. 2, and the calibration measurement data for 30B product and 48Y feed and tails cylinders is shown in Fig. 3. The data was collected in a low-background environment, meaning there is very little scatter in the data. The PNEM measurement concept was first presented at the 2010 Institute of Nuclear Materials Management (INMM) Annual Meeting. The physics design and uncertainty analysis were presented at the 2010 International Atomic Energy Agency (IAEA) Safeguards Symposium, and the mechanical and electrical designs and characterization measurements were published in the ESARDA Bulletin in 2011.« less

Investigating Uranium Concentrations in Groundwaters in the State of Idaho Using Kinetic Phosphorescence Analysis and Inductively Coupled Plasma Mass Spectrometry.

PubMed

Tkavadze, Levan; Dunker, Roy E; Brey, Richard R; Dudgeon, John

2016-11-01

The determination of uranium concentrations in natural water samples is of great interest due to the environmental consequences of this radionuclide. In this study, 380 groundwater samples from various locations within the state of Idaho were analyzed using two different techniques. The first method was Kinetic Phosphorescence Analysis (KPA), which gives the total uranium concentrations in water samples. The second analysis method was inductively coupled plasma mass spectrometry (ICP- MS). This method determines the total uranium concentration as well as the separate isotope concentrations of uranium. The U/U isotopic ratio was also measured for each sample to confirm that there was no depleted or enriched uranium present. The results were compared and mapped separately from each other. The study also found that in some areas of the state, natural uranium concentrations are relatively high.

78 FR 21100 - Low Enriched Uranium From France: Final Results of the Expedited Second Sunset Review of the...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-04-09

... received no response from the respondent interested parties, i.e., French uranium producers and exporters... Centralized Electronic Service System (IA ACCESS). IA ACCESS is available to registered users at http... the Internet at http://trade.gov/ia/ . The signed Decision Memorandum and electronic versions of the...

A wet chemical method for the estimation of carbon in uranium carbides.

PubMed

Chandramouli, V; Yadav, R B; Rao, P R

1987-09-01

A wet chemical method for the estimation of carbon in uranium carbides has been developed, based on oxidation with a saturated solution of sodium dichromate in 9M sulphuric acid, absorption of the evolved carbon dioxide in a known excess of barium hydroxide solution, and titration of the excess of barium hydroxide with standard potassium hydrogen phthalate solution. The carbon content obtained is in good agreement with that obtained by combustion and titration.

Francis Perrin's 1939 Analysis of Uranium Criticality

NASA Astrophysics Data System (ADS)

Reed, Cameron

2012-03-01

In May 1939, French physicist Francis Perrin published the first numerical estimate of the fast-neutron critical mass of a uranium compound. While his estimate of about 40 metric tons (12 tons if tamped) pertained to uranium oxide of natural isotopic composition as opposed to the enriched uranium that would be required for a nuclear weapon, it is interesting to examine Perrin's physics and to explore the subsequent impact of his paper. In this presentation I will discuss Perrin's model, the likely provenance of his parameter values, and how his work compared to the approach taken by Robert Serber in his 1943 Los Alamos Primer.

Target and method for the production of fission product molybdenum-99

DOEpatents

Vandegrift, George F.; Vissers, Donald R.; Marshall, Simon L.; Varma, Ravi

1989-01-01

A target for the reduction of fission product Mo-99 is prepared from uranium of low U-235 enrichment by coating a structural support member with a preparatory coating of a substantially oxide-free substrate metal. Uranium metal is electrodeposited from a molten halide electrolytic bath onto a substrate metal. The electrodeposition is performed at a predetermined direct current rate or by using pulsed plating techniques which permit relaxation of accumulated uranium ion concentrations within the melt. Layers of as much as to 600 mg/cm.sup.2 of uranium can be prepared to provide a sufficient density to produce acceptable concentrations of fission product Mo-99.

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

Federal Register 2010, 2011, 2012, 2013, 2014

2012-01-09

... France: Initiation of Antidumping Duty Changed Circumstances Review AGENCY: Import Administration... (Department) is initiating a changed circumstances review of the antidumping duty order on low enriched... On December 5, 2011, AREVA requested that the Department initiate and conduct an expedited changed...

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

Code of Federal Regulations, 2011 CFR

2011-01-01

... enrichment facilities authorized to produce special nuclear material of low strategic significance. 74.33... NUCLEAR MATERIAL Special Nuclear Material of Low Strategic Significance § 74.33 Nuclear material control... strategic significance. (a) General performance objectives. Each licensee who is authorized by this chapter...

78 FR 65389 - United States Enrichment Corporation, Paducah Gaseous Diffusion Plant

Federal Register 2010, 2011, 2012, 2013, 2014

2013-10-31

..., USEC notified the NRC of its decision to permanently cease uranium enrichment activities at the PGDP... Accession Nos. ML13105A010 and ML13176A151, respectively. NRC's PDR: You may examine and purchase copies of... in Paducah, Kentucky, using the gaseous [[Page 65390

Observed Changes in As-Fabricated U-10Mo Monolithic Fuel Microstructures After Irradiation in the Advanced Test Reactor

NASA Astrophysics Data System (ADS)

Keiser, Dennis; Jue, Jan-Fong; Miller, Brandon; Gan, Jian; Robinson, Adam; Madden, James

2017-12-01

A low-enriched uranium U-10Mo monolithic nuclear fuel is being developed by the Material Management and Minimization Program, earlier known as the Reduced Enrichment for Research and Test Reactors Program, for utilization in research and test reactors around the world that currently use high-enriched uranium fuels. As part of this program, reactor experiments are being performed in the Advanced Test Reactor. It must be demonstrated that this fuel type exhibits mechanical integrity, geometric stability, and predictable behavior to high powers and high fission densities in order for it to be a viable fuel for qualification. This paper provides an overview of the microstructures observed at different regions of interest in fuel plates before and after irradiation for fuel samples that have been tested. These fuel plates were fabricated using laboratory-scale fabrication methods. Observations regarding how microstructural changes during irradiation may impact fuel performance are discussed.

Biogeochemical prospecting for uranium with conifers: results from the Midnite Mine area, Washington

USGS Publications Warehouse

Nash, J. Thomas; Ward, Frederick Norville

1977-01-01

The ash of needles, cones, and duff from Ponderosa pine (Pinus ponderosa Laws) growing near uranium deposits of the Midnite mine, Stevens County, Wash., contain as much as 200 parts per million (ppm) uranium. Needle samples containing more than 10 ppm uranium define zones that correlate well with known uranium deposits or dumps. Dispersion is as much as 300 m but generally is less. Background is about 1 ppm. Tree roots are judged to be sampling ore, low-grade uranium halo, or ground water to a depth of about 15 m. Uptake of uranium by Douglas fir (Pseudotsuga menziesii (Mirb.) Franco) needles appears to be about the same as by Ponderosa pine needles. Cones and duff are generally enriched in uranium relate to needles. Needles, cones, and duff are recommended as easily collected, uncomplicated sample media for geochemical surveys. Samples can be analyzed by standard methods and total cost per sample kept to about $6.

Characterization of uranium bearing material using x-ray fluorescence and direct gamma-rays measurement techniques

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mujaini, M., E-mail: madihah@uniten.edu.my; Chankow, N.; Yusoff, M. Z.

2016-01-22

Uranium ore can be easily detected due to various gamma-ray energies emitted from uranium daughters particularly from {sup 238}U daughters such as {sup 214}Bi, {sup 214}Pb and {sup 226}Ra. After uranium is extracted from uranium ore, only low energy gamma-rays emitted from {sup 235}U may be detected if the detector is placed in close contact to the specimen. In this research, identification and characterization of uranium bearing materials is experimentally investigated using direct measurement of gamma-rays from {sup 235}U in combination with the x-ray fluorescence (XRF) technique. Measurement of gamma-rays can be conducted by using high purity germanium (HPGe) detectormore » or cadmium telluride (CdTe) detector while a {sup 57}Coradioisotope-excited XRF spectrometer using CdTe detector is used for elemental analysis. The proposed technique was tested with various uranium bearing specimens containing natural, depleted and enriched uranium in both metallic and powder forms.« less

An unattended verification station for UF6 cylinders: Field trial findings

NASA Astrophysics Data System (ADS)

Smith, L. E.; Miller, K. A.; McDonald, B. S.; Webster, J. B.; Zalavadia, M. A.; Garner, J. R.; Stewart, S. L.; Branney, S. J.; Todd, L. C.; Deshmukh, N. S.; Nordquist, H. A.; Kulisek, J. A.; Swinhoe, M. T.

2017-12-01

In recent years, the International Atomic Energy Agency (IAEA) has pursued innovative techniques and an integrated suite of safeguards measures to address the verification challenges posed by the front end of the nuclear fuel cycle. Among the unattended instruments currently being explored by the IAEA is an Unattended Cylinder Verification Station (UCVS), which could provide automated, independent verification of the declared relative enrichment, 235U mass, total uranium mass, and identification for all declared uranium hexafluoride cylinders in a facility (e.g., uranium enrichment plants and fuel fabrication plants). Under the auspices of the United States and European Commission Support Programs to the IAEA, a project was undertaken to assess the technical and practical viability of the UCVS concept. The first phase of the UCVS viability study was centered on a long-term field trial of a prototype UCVS system at a fuel fabrication facility. A key outcome of the study was a quantitative performance evaluation of two nondestructive assay (NDA) methods being considered for inclusion in a UCVS: Hybrid Enrichment Verification Array (HEVA), and Passive Neutron Enrichment Meter (PNEM). This paper provides a description of the UCVS prototype design and an overview of the long-term field trial. Analysis results and interpretation are presented with a focus on the performance of PNEM and HEVA for the assay of over 200 "typical" Type 30B cylinders, and the viability of an "NDA Fingerprint" concept as a high-fidelity means to periodically verify that material diversion has not occurred.

Molecular Dynamics Simulation of Thermodynamic Properties in Uranium Dioxide

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Xiangyu; Wu, Bin; Gao, Fei

2014-03-01

In the present study, we investigated the thermodynamic properties of uranium dioxide (UO2) by molecular dynamics (MD) simulations. As for solid UO2, the lattice parameter, density, and enthalpy obtained by MD simulations were in good agreement with existing experimental data and previous theoretical predictions. The calculated thermal conductivities matched the experiment results at the midtemperature range but were underestimated at very low and very high temperatures. The calculation results of mean square displacement represented the stability of uranium at all temperatures and the high mobility of oxygen toward 3000 K. By fitting the diffusivity constant of oxygen with the Vogel-Fulcher-Tammanmore » law, we noticed a secondary phase transition near 2006.4 K, which can be identified as a ‘‘strong’’ to ‘‘fragile’’ supercooled liquid or glass phase transition in UO2. By fitting the oxygen diffusion constant with the Arrhenius equation, activation energies of 2.0 and 2.7 eV that we obtained were fairly close to the recommended values of 2.3 to 2.6 eV. Xiangyu Wang, Bin Wu, Fei Gao, Xin Li, Xin Sun, Mohammed A. Khaleel, Ademola V. Akinlalu and Li Liu« less

Irradiation of TZM: Uranium dioxide fuel pin at 1700 K

NASA Technical Reports Server (NTRS)

Mcdonald, G. E.

1973-01-01

A fuel pin clad with TZM and containing solid pellets of uranium dioxide was fission heated in a static helium-cooled capsule at a maximum surface temperature of 1700 K for approximately 1000 hr and to a total burnup of 2.0 percent of the uranium-235. The results of the postirradiation examination indicated: (1) A transverse, intergranular failure of the fuel pin occurred when the fuel pin reached 2.0-percent burnup. This corresponds to 1330 kW-hr/cu cm, where the volume is the sum of the fuel, clad, and void volumes in the fuel region. (2) The maximum swelling of the fuel pin was less than 1.5 percent on the fuel-pin diameter. (3) There was no visible interaction between the TZM clad and the UO2. (4) Irradiation at 1700 K produced a course-grained structure, with an average grain diameter of 0.02 centimeter and with some of the grains extending one-half of the thickness of the clad. (5) Below approximately 1500 K, the irradiation of the clad produced a moderately fine-grained structure, with an average grain diameter of 0.004 centimeter.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stillman, J. A.; Feldman, E. E.; Jaluvka, D.

This report contains the results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members in the Research and Test Reactor Department at the Argonne National Laboratory (ANL) and the MURR Facility. MURR LEU conversion is part of an overall effort to develop and qualify high-density fuel within the U.S. High Performance Research Reactor Conversion (USHPRR) program conducted by the U.S. Department of Energy National Nuclearmore » Security Administration’s Office of Material Management and Minimization (M 3).« less

ANL progress on the cooperation with CNEA for the Mo-99 production : base-side digestion process.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gelis, A. V.; Quigley, K. J.; Aase, S. B.

2004-01-01

Conversion from high-enriched uranium (HEU) to low-enriched uranium (LEU) targets for the Mo-99 production requires certain modifications of the target design, the digestion and the purification processes. ANL is assisting the Argentine Comision Nacional de Energia Atomica (CNEA) to overcome all the concerns caused by the conversion to LEU foil targets. A new digester with stirring system has been successfully applied for the digestion of the low burn-up U foil targets in KMnO4 alkaline media. In this paper, we report the progress on the development of the digestion procedure with stirring focusing on the minimization of the liquid radioactive waste.

Portsmouth annual environmental report for 2003, Piketon, Ohio

DOE Office of Scientific and Technical Information (OSTI.GOV)

none, none

2004-11-30

The Portsmouth & Gaseous Diffusion Plant (PORTS) is located on a 5.8-square-mile site in a rural area of Pike County, Ohio. U.S. Department of Energy (DOE) activities at PORTS include environmental restoration, waste 'management, and long-term'stewardship of nonleased facilities: Production facilities for the separation of uranium isotopes are leased to the United States Enrichment Corporation (USEC), but most activities associated with the uranium enrichment process ceased in 2001. USEC activities are not covered by this document, with the exception of some environmental compliance information provided in Chap. 2 and radiological and non-radiological environmental monitoring program information discussed in Chaps. 4more » and 5.« less

Infrared Lasers in Chemistry.

ERIC Educational Resources Information Center

John, Phillip

1982-01-01

Selected infrared laser chemistry topics are discussed including carbon dioxide lasers, infrared quanta and molecules, laser-induced chemistry, structural isomerization (laser purification, sensitized reactions, and dielectric breakdown), and fundamental principles of laser isotope separation, focusing on uranium isotope separation. (JN)

Synthesis and characterization of surrogate nuclear explosion debris: urban glass matrix

DOE PAGES

Campbell, Keri; Judge, Elizabeth J.; Dirmyer, Matthew R.; ...

2017-07-26

Surrogate nuclear explosive debris was synthesized and characterized for major, minor, and trace elemental composition as well as uranium isotopics. The samples consisted of an urban glass matrix, equal masses soda lime and cement, doped with 500 ppm uranium with varying enrichments. The surface and cross section morphology were measured with SEM, and the major elemental composition was determined by XPS. LA-ICP-MS was used to measure the uranium isotopic abundance comparing different sampling techniques. Furthermore, the results provide an example of the utility of LA-ICP-MS for forensics applications.

Pena blanca natural analogue project: summary of activities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Levy, Schon S; Goldstein, Steven J; Abdel - Fattah, Amr I

2010-12-08

The inactive Nopal I uranium mine in silicic tuff north of Chihuahua City, Chihuahua, Mexico, was studied as a natural analogue for an underground nuclear-waste repository in the unsaturated zone. Site stratigraphy was confirmed from new drill core. Datafrom site studies include chemical and isotopic compositions of saturated- and unsaturated-zone waters. A partial geochronology of uranium enrichment and mineralization was established. Evidence pertinent to uranium-series transport in the soil zone and changing redox conditions was collected. The investigations contributed to preliminary, scoping-level performance assessment modeling.

[The risks of out of area missions: depleted uranium].

PubMed

Ciprani, F; Moroni, M

2006-01-01

Depleted uranium (DU), a waste product of uranium enrichment, has several civilian and military applications. It was used as armor-piercing ammunition in international conflicts and was claimed to contribute to health problems, known as the Gulf War Syndrome and recently as the Balkan Syndrome. Leukaemia/Limphoma cases among UN soldiers in the Balkans have been related hypothetically to exposure to DU. The investigations published in the scientific literature give no support for this hypothesis. However future follow-up is necessary for evaluation of long-term risk.

METHOD OF DISSOLVING REFRACTORY ALLOYS

DOEpatents

Helton, D.M.; Savolainen, J.K.

1963-04-23

This patent relates to the dissolution of alloys of uranium with zirconium, thorium, molybdenum, or niobium. The alloy is contacted with an anhydrous solution of mercuric chloride in a low-molecular-weight monohydric alcohol to produce a mercury-containing alcohol slurry. The slurry is then converted to an aqueous system by adding water and driving off the alcohol. The resulting aqueous slurry is electrolyzed in the presence of a mercury cathode to remove the mercury and produce a uranium-bearing aqueous solution. This process is useful for dissolving irradiated nuclear reactor fuels for radiochemical reprocessing by solvent extraction. In addition, zirconium-alloy cladding is selectively removed from uranium dioxide fuel compacts by this means. (AEC)

Mathematical simulation of the amplification of 1790-nm laser radiation in a nuclear-excited He - Ar plasma containing nanoclusters of uranium compounds

NASA Astrophysics Data System (ADS)

Kosarev, V. A.; Kuznetsova, E. E.

2014-02-01

The possibility of applying dusty active media in nuclearpumped lasers has been considered. The amplification of 1790-nm radiation in a nuclear-excited dusty He - Ar plasma is studied by mathematical simulation. The influence of nanoclusters on the component composition of the medium and the kinetics of the processes occurring in it is analysed using a specially developed kinetic model, including 72 components and more than 400 reactions. An analysis of the results indicates that amplification can in principle be implemented in an active laser He - Ar medium containing 10-nm nanoclusters of metallic uranium and uranium dioxide.

Different pattern of brain pro-/anti-oxidant activity between depleted and enriched uranium in chronically exposed rats.

PubMed

Lestaevel, P; Romero, E; Dhieux, B; Ben Soussan, H; Berradi, H; Dublineau, I; Voisin, P; Gourmelon, P

2009-04-05

Uranium is not only a heavy metal but also an alpha particle emitter. The main toxicity of uranium is expected to be due to chemiotoxicity rather than to radiotoxicity. Some studies have demonstrated that uranium induced some neurological disturbances, but without clear explanations. A possible mechanism of this neurotoxicity could be the oxidative stress induced by reactive oxygen species imbalance. The aim of the present study was to determine whether a chronic ingestion of uranium induced anti-oxidative defence mechanisms in the brain of rats. Rats received depleted (DU) or 4% enriched (EU) uranyl nitrate in the drinking water at 2mg(-1)kg(-1)day(-1) for 9 months. Cerebral cortex analyses were made by measuring mRNA and protein levels and enzymatic activities. Lipid peroxidation, an oxidative stress marker, was significantly enhanced after EU exposure, but not after DU. The gene expression or activity of the main antioxidant enzymes, i.e. superoxide dismutase (SOD), catalase (CAT) and glutathione peroxidase (GPx), increased significantly after chronic exposure to DU. On the contrary, oral EU administration induced a decrease of these antioxidant enzymes. The NO-ergic pathway was almost not perturbed by DU or EU exposure. Finally, DU exposure increased significantly the transporters (Divalent-Metal-Transporter1; DMT1), the storage molecule (ferritin) and the ferroxidase enzyme (ceruloplasmin), but not EU. These results illustrate that oxidative stress plays a key role in the mechanism of uranium neurotoxicity. They showed that chronic exposure to DU, but not EU, seems to induce an increase of several antioxidant agents in order to counteract the oxidative stress. Finally, these results demonstrate the importance of the double toxicity, chemical and radiological, of uranium.

Environmental site description for a Uranium Atomic Vapor Laser Isotope Separation (U-AVLIS) production plant at the Paducah Gaseous Diffusion Plant site

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marmer, G.J.; Dunn, C.P.; Moeller, K.L.

Uranium enrichment in the United States has utilized a diffusion process to preferentially enrich the U-235 isotope in the uranium product. The U-AVLIS process is based on electrostatic extraction of photoionized U-235 atoms from an atomic vapor stream created by electron-beam vaporization of uranium metal alloy. The U-235 atoms are ionized when precisely tuned laser light -- of appropriate power, spectral, and temporal characteristics -- illuminates the uranium vapor and selectively photoionizes the U-235 isotope. A programmatic document for use in screening DOE site to locate a U-AVLIS production plant was developed and implemented in two parts. The first partmore » consisted of a series of screening analyses, based on exclusionary and other criteria, that identified a reasonable number of candidate sites. These sites were subjected to a more rigorous and detailed comparative analysis for the purpose of developing a short list of reasonable alternative sites for later environmental examination. This environmental site description (ESD) provides a detailed description of the PGDP site and vicinity suitable for use in an environmental impact statement (EIS). The report is based on existing literature, data collected at the site, and information collected by Argonne National Laboratory (ANL) staff during a site visit. 65 refs., 15 tabs.« less

Preliminary investigation of the elemental variation and diagenesis of a tabular uranium deposit, La Sal Mine, San Juan County, Utah

USGS Publications Warehouse

Brooks, Robert A.; Campbell, John A.

1976-01-01

Ore in the La Sal mine, San Juan County, Utah, occurs as a typical tabular-type uranium deposit of the-Colorado Plateau. Uranium-vanadium occurs in the Salt Wash Member of the Jurassic Morrison Formation. Chemical and petrographic analyses were used to determine elemental variation and diagenetic aspects across the orebody. Vanadium is concentrated in the dark clay matrix, which constitutes visible ore. Uranium content is greater above the vanadium zone. Calcium, carbonate carbon, and lead show greater than fifty-fold increase across the ore zone, whereas copper and organic carbon show only a several-fold increase. Large molybdenum concentrations are present in and above the tabular layer, and large selenium concentrations occur below the uranium zone within the richest vanadium zone. Iron is enriched in the vanadium horizon. Chromium is depleted from above the ore and strongly enriched below. Elements that vary directly with the vanadium content include magnesium, iron, selenium, zirconium, strontium, titanium, lead, boron, yttrium, and scandium. The diagenetic sequence is as follows: (1) formation of secondary quartz overgrowths as cement; (2) infilling and lining of remaining pores with amber opaline material; (3) formation of vanadium-rich clay matrix, which has replaced overgrowths as well as quartz grains; (4) replacement of overgrowths and detrital grains by calcite; (5) infilling of pores with barite and the introduction of pyrite and marcasite.

WNA's worldwide overview on front-end nuclear fuel cycle growth and health, safety and environmental issues.

PubMed

Saint-Pierre, Sylvain; Kidd, Steve

2011-01-01

This paper presents the WNA's worldwide nuclear industry overview on the anticipated growth of the front-end nuclear fuel cycle from uranium mining to conversion and enrichment, and on the related key health, safety, and environmental (HSE) issues and challenges. It also puts an emphasis on uranium mining in new producing countries with insufficiently developed regulatory regimes that pose greater HSE concerns. It introduces the new WNA policy on uranium mining: Sustaining Global Best Practices in Uranium Mining and Processing-Principles for Managing Radiation, Health and Safety and the Environment, which is an outgrowth of an International Atomic Energy Agency (IAEA) cooperation project that closely involved industry and governmental experts in uranium mining from around the world. Copyright © 2010 Health Physics Society

Uranium and other contaminants in hair from the parents of children with congenital anomalies in Fallujah, Iraq

PubMed Central

2011-01-01

Background Recent reports have drawn attention to increases in congenital birth anomalies and cancer in Fallujah Iraq blamed on teratogenic, genetic and genomic stress thought to result from depleted Uranium contamination following the battles in the town in 2004. Contamination of the parents of the children and of the environment by Uranium and other elements was investigated using Inductively Coupled Plasma Mass Spectrometry. Hair samples from 25 fathers and mothers of children diagnosed with congenital anomalies were analysed for Uranium and 51 other elements. Mean ages of the parents was: fathers 29.6 (SD 6.2); mothers: 27.3 (SD 6.8). For a sub-group of 6 women, long locks of hair were analysed for Uranium along the length of the hair to obtain information about historic exposures. Samples of soil and water were also analysed and Uranium isotope ratios determined. Results Levels of Ca, Mg, Co, Fe, Mn, V, Zn, Sr, Al, Ba, Bi, Ga, Pb, Hg, Pd and U (for mothers only) were significantly higher than published mean levels in an uncontaminated population in Sweden. In high excess were Ca, Mg, Sr, Al, Bi and Hg. Of these only Hg can be considered as a possible cause of congenital anomaly. Mean levels for Uranium were 0.16 ppm (SD: 0.11) range 0.02 to 0.4, higher in mothers (0.18 ppm SD 0.09) than fathers (0.11 ppm; SD 0.13). The highly unusual non-normal Fallujah distribution mean was significantly higher than literature results for a control population Southern Israel (0.062 ppm) and a non-parametric test (Mann Whitney-Wilcoxon) gave p = 0.016 for this comparison of the distribution. Mean levels in Fallujah were also much higher than the mean of measurements reported from Japan, Brazil, Sweden and Slovenia (0.04 ppm SD 0.02). Soil samples show low concentrations with a mean of 0.76 ppm (SD 0.42) and range 0.1-1.5 ppm; (N = 18). However it may be consistent with levels in drinking water (2.28 μgL-1) which had similar levels to water from wells (2.72 μgL-1) and the river Euphrates (2.24 μgL-1). In a separate study of a sub group of mothers with long hair to investigate historic Uranium excretion the results suggested that levels were much higher in the past. Uranium traces detected in the soil samples and the hair showed slightly enriched isotopic signatures for hair U238/U235 = (135.16 SD 1.45) compared with the natural ratio of 137.88. Soil sample Uranium isotope ratios were determined after extraction and concentration of the Uranium by ion exchange. Results showed statistically significant presence of enriched Uranium with a mean of 129 with SD5.9 (for this determination, the natural Uranium 95% CI was 132.1 < Ratio < 144.1). Conclusions Whilst caution must be exercised about ruling out other possibilities, because none of the elements found in excess are reported to cause congenital diseases and cancer except Uranium, these findings suggest the enriched Uranium exposure is either a primary cause or related to the cause of the congenital anomaly and cancer increases. Questions are thus raised about the characteristics and composition of weapons now being deployed in modern battlefields PMID:21888647

Diversification in the Supply Chain of (99)Mo Ensures a Future for (99m)Tc.

PubMed

Cutler, Cathy S; Schwarz, Sally W

2014-07-01

The uncertain availability of (99m)Tc has become a concern for nuclear medicine departments across the globe. An issue for the United States is that currently it is dependent on a supply of (99m)Tc (from (99)Mo) that is derived solely by production outside the United States. Since the United States uses half the world's (99)Mo production, the U.S. (99)Mo supply chain would be greatly enhanced if a producer were located within the United States. The fragility of the old (99)Mo supply chain is being addressed as new facilities are constructed and new processes are developed to produce (99)Mo without highly enriched uranium. The conversion to low-enriched uranium is necessary to minimize the potential misuse of highly enriched uranium in the world for nonpeaceful means. New production facilities, new methods for the production of (99)Mo, and a new generator elution system for the supply of (99m)Tc are currently being pursued. The progress made in all these areas will be discussed, as they all highlight the need to embrace diversity to ensure that we have a robust and reliable supply of (99m)Tc in the future. © 2014 by the Society of Nuclear Medicine and Molecular Imaging, Inc.

High-Fidelity Modelng and Simulation for a High Flux Isotope Reactor Low-Enriched Uranium Core Design

DOE PAGES

Betzler, Benjamin R.; Chandler, David; Davidson, Eva E.; ...

2017-05-08

A high-fidelity model of the High Flux Isotope Reactor (HFIR) with a low-enriched uranium (LEU) fuel design and a representative experiment loading has been developed to serve as a new reference model for LEU conversion studies. With the exception of the fuel elements, this HFIR LEU model is completely consistent with the current highly enriched uranium HFIR model. Results obtained with the new LEU model provide a baseline for analysis of alternate LEU fuel designs and further optimization studies. The newly developed HFIR LEU model has an explicit representation of the HFIR-specific involute fuel plate geometry, including the within-plate fuelmore » meat contouring, and a detailed geometry model of the fuel element side plates. Such high-fidelity models are necessary to accurately account for the self-shielding from 238U and the depletion of absorber materials present in the side plates. In addition, a method was developed to account for fuel swelling in the high-density LEU fuel plates during the depletion simulation. In conclusion, calculated time-dependent metrics for the HFIR LEU model include fission rate and cumulative fission density distributions, flux and reaction rates for relevant experiment locations, point kinetics data, and reactivity coefficients.« less

Air Shipment of Highly Enriched Uranium Spent Nuclear Fuel from Romania

DOE Office of Scientific and Technical Information (OSTI.GOV)

K. J. Allen; I. Bolshinsky; L. L. Biro

2010-07-01

Romania safely air shipped 23.7 kilograms of Russian origin highly enriched uranium (HEU) spent nuclear fuel from the VVR S research reactor at Magurele, Romania, to the Russian Federation in June 2009. This was the world’s first air shipment of spent nuclear fuel transported in a Type B(U) cask under existing international laws without special exceptions for the air transport licenses. This shipment was coordinated by the Russian Research Reactor Fuel Return Program (RRRFR), part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), in cooperation with the Romania National Commission for Nuclear Activities Control (CNCAN), the Horiamore » 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.« less

High-Fidelity Modelng and Simulation for a High Flux Isotope Reactor Low-Enriched Uranium Core Design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Betzler, Benjamin R.; Chandler, David; Davidson, Eva E.

A high-fidelity model of the High Flux Isotope Reactor (HFIR) with a low-enriched uranium (LEU) fuel design and a representative experiment loading has been developed to serve as a new reference model for LEU conversion studies. With the exception of the fuel elements, this HFIR LEU model is completely consistent with the current highly enriched uranium HFIR model. Results obtained with the new LEU model provide a baseline for analysis of alternate LEU fuel designs and further optimization studies. The newly developed HFIR LEU model has an explicit representation of the HFIR-specific involute fuel plate geometry, including the within-plate fuelmore » meat contouring, and a detailed geometry model of the fuel element side plates. Such high-fidelity models are necessary to accurately account for the self-shielding from 238U and the depletion of absorber materials present in the side plates. In addition, a method was developed to account for fuel swelling in the high-density LEU fuel plates during the depletion simulation. In conclusion, calculated time-dependent metrics for the HFIR LEU model include fission rate and cumulative fission density distributions, flux and reaction rates for relevant experiment locations, point kinetics data, and reactivity coefficients.« less

A cellular automaton method to simulate the microstructure and evolution of low-enriched uranium (LEU) U-Mo/Al dispersion type fuel plates

NASA Astrophysics Data System (ADS)

Drera, Saleem S.; Hofman, Gerard L.; Kee, Robert J.; King, Jeffrey C.

2014-10-01

Low-enriched uranium (LEU) fuel plates for high power materials test reactors (MTR) are composed of nominally spherical uranium-molybdenum (U-Mo) particles within an aluminum matrix. Fresh U-Mo particles typically range between 10 and 100 μm in diameter, with particle volume fractions up to 50%. As the fuel ages, reaction-diffusion processes cause the formation and growth of interaction layers that surround the fuel particles. The growth rate depends upon the temperature and radiation environment. The cellular automaton algorithm described in this paper can synthesize realistic random fuel-particle structures and simulate the growth of the intermetallic interaction layers. Examples in the present paper pack approximately 1000 particles into three-dimensional rectangular fuel structures that are approximately 1 mm on each side. The computational approach is designed to yield synthetic microstructures consistent with images from actual fuel plates and is validated by comparison with empirical data on actual fuel plates.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

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

Minimum Nuclear Deterrence Postures in South Asia: An Overview

DTIC Science & Technology

2001-10-01

states in May 1998, India and Pakistan both espoused nuclear restraint. Their senior officials soon embraced the language of "minimum credible...Air Force and Army. India’s longer-range nuclear-capable missiles such as the Agni, however, are still in the research and development process under...explained in Appendix A, Pakistan continued between 1991 and 1998 to enrich uranium to low- enriched (LEU) levels. Since enrichment is an iterative process

Pyroprocessing of Fast Flux Test Facility Nuclear Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

B.R. Westphal; G.L. Fredrickson; G.G. Galbreth

Used nuclear fuel from the Fast Flux Test Facility (FFTF) was recently transferred to the Idaho National Laboratory and processed by pyroprocessing in the Fuel Conditioning Facility. Approximately 213 kg of uranium from sodium-bonded metallic FFTF fuel was processed over a one year period with the equipment previously used for the processing of EBR-II used fuel. The peak burnup of the FFTF fuel ranged from 10 to 15 atom% for the 900+ chopped elements processed. Fifteen low-enriched uranium ingots were cast following the electrorefining and distillation operations to recover approximately 192 kg of uranium. A material balance on the primarymore » fuel constituents, uranium and zirconium, during the FFTF campaign will be presented along with a brief description of operating parameters. Recoverable uranium during the pyroprocessing of FFTF nuclear fuel was greater than 95% while the purity of the final electrorefined uranium products exceeded 99%.« less

Pyroprocessing of fast flux test facility nuclear fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Westphal, B.R.; Wurth, L.A.; Fredrickson, G.L.

Used nuclear fuel from the Fast Flux Test Facility (FFTF) was recently transferred to the Idaho National Laboratory and processed by pyroprocessing in the Fuel Conditioning Facility. Approximately 213 kg of uranium from sodium-bonded metallic FFTF fuel was processed over a one year period with the equipment previously used for the processing of EBR-II used fuel. The peak burnup of the FFTF fuel ranged from 10 to 15 atom% for the 900+ chopped elements processed. Fifteen low-enriched uranium ingots were cast following the electrorefining and distillation operations to recover approximately 192 kg of uranium. A material balance on the primarymore » fuel constituents, uranium and zirconium, during the FFTF campaign will be presented along with a brief description of operating parameters. Recoverable uranium during the pyroprocessing of FFTF nuclear fuel was greater than 95% while the purity of the final electro-refined uranium products exceeded 99%. (authors)« less

Identifying anthropogenic uranium compounds using soft X-ray near-edge absorption spectroscopy

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ward, Jesse D.; Bowden, Mark; Tom Resch, C.

2017-01-01

Uranium ores mined for industrial use are typically acid-leached to produce yellowcake and then converted into uranium halides for enrichment and purification. These anthropogenic chemical forms of uranium are distinct from their mineral counterparts. The purpose of this study is to use soft X-ray absorption spectroscopy to characterize several common anthropogenic uranium compounds important to the nuclear fuel cycle. Non-destructive chemical analyses of these compounds is important for process and environmental monitoring and X-ray absorption techniques have several advantages in this regard, including element-specificity, chemical sensitivity, and high spectral resolution. Oxygen K-edge spectra were collected for uranyl nitrate, uranyl fluoride,more » and uranyl chloride, and fluorine K-edge spectra were collected for uranyl fluoride and uranium tetrafluoride. Interpretation of the data is aided by comparisons to calculated spectra. These compounds have unique spectral signatures that can be used to identify unknown samples.« less

Production of fissioning uranium plasma to approximate gas-core reactor conditions

NASA Technical Reports Server (NTRS)

Lee, J. H.; Mcfarland, D. R.; Hohl, F.; Kim, K. H.

1974-01-01

The intense burst of neutrons from the d-d reaction in a plasma-focus apparatus is exploited to produce a fissioning uranium plasma. The plasma-focus apparatus consists of a pair of coaxial electrodes and is energized by a 25 kJ capacitor bank. A 15-g rod of 93% enriched U-235 is placed in the end of the center electrode where an intense electron beam impinges during the plasma-focus formation. The resulting uranium plasma is heated to about 5 eV. Fission reactions are induced in the uranium plasma by neutrons from the d-d reaction which were moderated by the polyethylene walls. The fission yield is determined by evaluating the gamma peaks of I-134, Cs-138, and other fission products, and it is found that more than 1,000,000 fissions are induced in the uranium for each focus formation, with at least 1% of these occurring in the uranium plasma.

Response of vegetation to carbon dioxide. Growth, yield and plant water relationships in sweet potatoes in response to carbon dioxide enrichment 1986

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

1998-08-01

In the summer of 1985, under the joint program of US Department of Energy, Carbon Dioxide Division, and Tuskegee University, experiments were conducted to study growth, yield, photosynthesis and plant water relationships in sweet potato plants growth in an enriched CO{sub 2} environment. The main experiment utilized open top chambers to study the effects of CO{sub 2} and soil moisture on growth, yield and photosynthesis of field-grown plants. In addition, potted plants in open top chambers were utilized in a study of the effects of different CO{sub 2} concentrations on growth pattern, relative growth rate, net assimilation rate and biomassmore » increment at different stages of development. The interaction effects of enriched CO{sub 2} and water stress on biomass production, yield, xylem potential, and stomatal conductance were also investigated. The overall results of the various studies are described.« less

Lessons-Learned from D and D Activities at the Five Gaseous Diffusion Buildings (K-25, K- 27, K-29, K-31 and K-33) East Tennessee Technology Park, Oak Ridge, TN - 13574

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kopotic, James D.; Ferri, Mark S.; Buttram, Claude

The East Tennessee Technology Park (ETTP) is the site of five former gaseous diffusion plant (GDP) process buildings that were used to enrich uranium from 1945 to 1985. The process equipment in the original two buildings (K-25 and K-27) was used for the production of highly enriched uranium (HEU), while that in the three later buildings (K-29, K-31 and K-33) produced low enriched uranium (LEU). Equipment was contaminated primarily with uranium and to a lesser extent technetium (Tc). Decommissioning of the GDP process buildings has presented several unique challenges and produced many lessons-learned. Among these is the importance of good,more » up-front characterization in developing the best demolition approach. Also, chemical cleaning of process gas equipment and piping (PGE) prior to shutdown should be considered to minimize the amount of hold-up material that must be removed by demolition crews. Another lesson learned is to maintain shutdown buildings in a dry state to minimize structural degradation which can significantly complicate characterization, deactivation and demolition efforts. Perhaps the most important lesson learned is that decommissioning GDP process buildings is first and foremost a waste logistics challenge. Innovative solutions are required to effectively manage the sheer volume of waste generated from decontamination and demolition (D and D) of these enormous facilities. Finally, close coordination with Security is mandatory to effectively manage Special Nuclear Material (SNM) and classified equipment issues. (authors)« less

Ion Mobility Spectrometer Field Test

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, Nicholas; McLain, Derek; Steeb, Jennifer

The Morpho Saffran Itemizer 4DX Ion Mobility Spectrometer previously used to detect uranium signatures in FY16 was used at the former New Brunswick Facility, a past uranium facility located on site at Argonne National Laboratory. This facility was chosen in an attempt to detect safeguards relevant signatures and has a history of processing uranium at various enrichments, chemical forms, and purities; various chemicals such as nitric acid, uranium fluorides, phosphates and metals are present at various levels. Several laboratories were sampled for signatures of nuclear activities around the laboratory. All of the surfaces that were surveyed were below background levelsmore » of the radioanalytical instrumentation and determined to be radiologically clean.« less

Target and method for the production of fission product molybdenum-99

DOEpatents

Vandegrift, G.F.; Vissers, D.R.; Marshall, S.L.; Varma, R.

1987-10-26

A target for the reduction of fission product Mo-99 is prepared from uranium of low U-235 enrichment by coating a structural support member with a preparatory coating of a substantially oxide-free substrate metal. Uranium metal is electrodeposited from a molten halide electrolytic bath onto a substrate metal. The electrodeposition is performed at a predetermined direct current rate or by using pulsed plating techniques which permit relaxation of accumulated uranium ion concentrations within the melt. Layers of as much as to 600 mg/cm/sup 2/ of uranium can be prepared to provide a sufficient density to produce acceptable concentrations of fission product Mo-99. 2 figs.

REDUCTION OF INORGANIC COMPOUNDS WITH MOLECULAR HYDROGEN BY MICROCOCCUS LACTILYTICUS I.

PubMed Central

Woolfolk, C. A.; Whiteley, H. R.

1962-01-01

Woolfolk, C. A. (University of Washington, Seattle) and H. R. Whiteley. Reduction of inorganic compounds with molecular hydrogen by Micrococcus lactilyticus. I. Stoichiometry with compounds of arsenic, selenium, tellurium, transition and other elements. J. Bacteriol. 84:647–658. 1962.—Extracts of Micrococcus lactilyticus (Veillonella alcalescens) oxidize molecular hydrogen at the expense of certain compounds of arsenic, bismuth, selenium, tellurium, lead, thallium, vanadium, manganese, iron, copper, molybdenum, tungsten, osmium, ruthenium, gold, silver, and uranium, as well as molecular oxygen. Chemical and manometric data indicate that the following reductions are essentially quantitative: arsenate to arsenite, pentavalent and trivalent bismuth to the free element, selenite via elemental selenium to selenide, tellurate and tellurite to tellurium, lead dioxide and manganese dioxide to the divalent state, ferric to ferrous iron, osmium tetroxide to osmate ion, osmium dioxide and trivalent osmium to the metal, uranyl uranium to the tetravalent state, vanadate to the level of vanadyl, and polymolybdate ions to molybdenum blues with an average valence for molybdenum of +5. The results of a study of certain other hydrogenase-containing bacteria with respect to their ability to carry out some of the same reactions are also presented. PMID:14001842

Analysis of protein phosphorylation by monolithic extraction columns based on poly(divinylbenzene) containing embedded titanium dioxide and zirconium dioxide nano-powders.

PubMed

Rainer, Matthias; Sonderegger, Harald; Bakry, Rania; Huck, Christian W; Morandell, Sandra; Huber, Lukas A; Gjerde, Douglas T; Bonn, Günther K

2008-11-01

The potential of an organic monolith with incorporated titanium dioxide (TiO(2)) and zirconium dioxide (ZrO(2)) nanoparticles was evaluated for the selective enrichment of phosphorylated peptides from tryptic digests. A pipette tip was fitted with a monolith based on divinylbenzene (DVB) of highly porous structure, which allows sample to pass through the monolithic bed. The enrichment of phosphopeptides was enhanced by increasing the pipetting cycles during the sample preparation and a higher recovery could be achieved with adequate buffer systems. A complete automated process was developed for enrichment of phosphopeptides leading to high reproducibility and resulting in a robust method designed to minimize analytical variance while providing high sensitivity at high sample throughput. The effect of particle size on the selectivity of phosphopeptides was investigated by comparative studies with nano- and microscale TiO(2) and ZrO(2) powders. Eleven phosphopeptides from alpha-casein digest could be recovered by an optimized mixture of microscale TiO(2)/ZrO(2) particles, whereas nine additional phosphopeptides could be retained by the same mixture of nano-structured material. When compared to conventional immobilized metal-ion affinity chromatography and commercial phosphorylation-enrichment kits, higher selectivity was observed in case of self fabricated tips. About 20 phosphopeptides could be retained from alpha-casein and five from beta-casein digests by using TiO(2) and ZrO(2) based extraction tips. Further selectivity for phosphopeptides was demonstrated by enriching a digest of in vitro phosphorylated extracellular signal regulated kinase 1 (ERK1). Two phosphorylated peptides of ERK1 could be identified by MALDI-MS/MS measurements and a following MASCOT database search.

The prospect of uranium nitride (UN) and mixed nitride fuel (UN-PuN) for pressurized water reactor

NASA Astrophysics Data System (ADS)

Syarifah, Ratna Dewi; Suud, Zaki

2015-09-01

Design study of small Pressurized Water Reactors (PWRs) core loaded with uranium nitride fuel (UN) and mixed nitride fuel (UN-PuN), Pa-231 as burnable poison, and Americium has been performed. Pa-231 known as actinide material, have large capture cross section and can be converted into fissile material that can be utilized to reduce excess reactivity. Americium is one of minor actinides with long half life. The objective of adding americium is to decrease nuclear spent fuel in the world. The neutronic analysis results show that mixed nitride fuel have k-inf greater than uranium nitride fuel. It is caused by the addition of Pu-239 in mixed nitride fuel. In fuel fraction analysis, for uranium nitride fuel, the optimum volume fractions are 45% fuel fraction, 10% cladding and 45% moderator. In case of UN-PuN fuel, the optimum volume fractions are 30% fuel fraction, 10% cladding and 60% coolant/ moderator. The addition of Pa-231 as burnable poison for UN fuel, enrichment U-235 5%, with Pa-231 1.6% has k-inf more than one and excess reactivity of 14.45%. And for mixed nitride fuel, the lowest value of reactivity swing is when enrichment (U-235+Pu) 8% with Pa-231 0.4%, the excess reactivity value 13,76%. The fuel pin analyze for the addition of Americium, the excess reactivity value is lower than before, because Americium absorb the neutron. For UN fuel, enrichment U-235 8%, Pa-231 1.6% and Am 0.5%, the excess reactivity is 4.86%. And for mixed nitride fuel, when enrichment (U-235+Pu) 13%, Pa-231 0.4% and Am 0.1%, the excess reactivity is 11.94%. For core configuration, it is better to use heterogeneous than homogeneous core configuration, because the radial power distribution is better.

Trace elements and Pb isotopes in soils and sediments impacted by uranium mining.

PubMed

Cuvier, A; Pourcelot, L; Probst, A; Prunier, J; Le Roux, G

2016-10-01

The purpose of this study is to evaluate the contamination in As, Ba, Co, Cu, Mn, Ni, Sr, V, Zn and REE, in a high uranium activity (up to 21,000Bq∙kg(-1)) area, downstream of a former uranium mine. Different geochemical proxies like enrichment factor and fractions from a sequential extraction procedure are used to evaluate the level of contamination, the mobility and the availability of the potential contaminants. Pb isotope ratios are determined in the total samples and in the sequential leachates to identify the sources of the contaminants and to determine the mobility of radiogenic Pb in the context of uranium mining. In spite of the large uranium contamination measured in the soils and the sediments (EF≫40), trace element contamination is low to moderate (2

Dose-response relationships between internally-deposited uranium and select health outcomes in gaseous diffusion plant workers, 1948-2011.

PubMed

Yiin, James H; Anderson, Jeri L; Bertke, Stephen J; Tollerud, David J

2018-05-09

To examine dose-response relationships between internal uranium exposures and select outcomes among a cohort of uranium enrichment workers. Cox regression was conducted to examine associations between selected health outcomes and cumulative internal uranium with consideration for external ionizing radiation, work-related medical X-rays and contaminant radionuclides technetium ( 99 Tc) and plutonium ( 239 Pu) as potential confounders. Elevated and monotonically increasing mortality risks were observed for kidney cancer, chronic renal diseases, and multiple myeloma, and the association with internal uranium absorbed organ dose was statistically significant for multiple myeloma. Adjustment for potential confounders had minimal impact on the risk estimates. Kidney cancer, chronic renal disease, and multiple myeloma mortality risks were elevated with increasing internal uranium absorbed organ dose. The findings add to evidence of an association between internal exposure to uranium and cancer. Future investigation includes a study of cancer incidence in this cohort. © 2018 Wiley Periodicals, Inc.

Study of Chemical Changes in Uranium Oxyfluoride Particles Progress Report March - October 2009

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kips, R; Kristo, M; Hutcheon, I

2009-11-22

Nuclear forensics relies on the analysis of certain sample characteristics to determine the origin and history of a nuclear material. In the specific case of uranium enrichment facilities, it is the release of trace amounts of uranium hexafluoride (UF{sub 6}) gas - used for the enrichment of uranium - that leaves a process-characteristic fingerprint. When UF{sub 6} gas interacts with atmospheric moisture, uranium oxyfluoride particles or particle agglomerates are formed with sizes ranging from several microns down to a few tens of nanometers. These particles are routinely collected by safeguards organizations, such as the International Atomic Energy Agency (IAEA), allowingmore » them to verify whether a facility is compliant with its declarations. Spectrometric analysis of uranium particles from UF{sub 6} hydrolysis has revealed the presence of both particles that contain fluorine, and particles that do not. It is therefore assumed that uranium oxyfluoride is unstable, and decomposes to form uranium oxide. Understanding the rate of fluorine loss in uranium oxyfluoride particles, and the parameters that control it, may therefore contribute to placing boundaries on the particle's exposure time in the environment. Expressly for the purpose of this study, we prepared a set of uranium oxyfluoride particles at the Institute for Reference Materials and Measurements (EU-JRC-IRMM) from a static release of UF{sub 6} in a humid atmosphere. The majority of the samples was stored in controlled temperature, humidity and lighting conditions. Single particles were characterized by a suite of micro-analytical techniques, including NanoSIMS, micro-Raman spectrometry (MRS), scanning (SEM) and transmission (TEM) electron microscopy, energy-dispersive X-ray spectrometry (EDX) and focused ion beam (FIB). The small particle size was found to be the main analytical challenge. The relative amount of fluorine, as well as the particle chemical composition and morphology were determined at different stages in the ageing process, and immediately after preparation. This report summarizes our most recent findings for each of the analytical techniques listed above, and provides an outlook on what remains to be resolved. Additional spectroscopic and mass spectrometric measurements were carried out at Pacific Northwest National Laboratory, but are not included in this summary.« less

China and Proliferation of Weapons of Mass Destruction and Missiles: Policy Issues

DTIC Science & Technology

2010-08-16

nuclear weapons facilities, while experts from China worked at a uranium mine at Saghand and a centrifuge facility (for uranium enrichment) near...brief interruptions.”85 84 Barbara Opall -Rome and Vago Muradian, “Bush Privately Lauds...confiscated a rare metal used to produce alloy steel (called vanadium) being smuggled to North Korea. In the same month, China’s NHI Shenyang Mining

Deploying Nuclear Detection Systems: A Proposed Strategy for Combating Nuclear Terrorism

DTIC Science & Technology

2007-07-01

lower cost than other gamma radiation detectors (if increased count rate is all one is looking for). Low cost makes plastic scintillation detectors...material, particularly enriched uranium and plutonium, the basic fuel for nuclear bombs. • Measures to strengthen international institutions to... uranium to specifications required for a nuclear weapon.1 This illicit shipment of centrifuges was part of an international nuclear materials

9. VIEW OF MOLTEN SALT BATH EQUIPMENT AND ROLLER PRESSES ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

9. VIEW OF MOLTEN SALT BATH EQUIPMENT AND ROLLER PRESSES BEING INSTALLED ON THE WEST SIDE (SIDE B) OF BUILDING 883. SIDE B OF BUILDING 883 WAS USED TO PROCESS ENRICHED URANIUM FROM 1957-66. (1/23/57) - Rocky Flats Plant, Uranium Rolling & Forming Operations, Southeast section of plant, southeast quadrant of intersection of Central Avenue & Eighth Street, Golden, Jefferson County, CO

HEU Holdup Measurements in the 321-M Draw Bench, Straightener, and Fluoroscope Components

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dewberry, R.A.

The Analytical Development Section of Savannah River Technology Center (SRTC) was requested by the Facilities Disposition Division (FDD) to determine the holdup of enriched uranium in the 321-M facility as part of an overall deactivation project of the facility. This report covers holdup measurements of uranium residue on the draw bench, straightener, and the fluoroscope components of the 321-M facility.

PROCESS FOR PRODUCING URANIUM HEXAFLUORIDE

DOEpatents

Fowler, R.D.

1957-10-22

A process for the production of uranium hexafluoride from the oxides of uranium is reported. In accordance with the method the higher oxides of uranium may be reduced to uranium dioxide (UO/sub 2/), the latter converted into uranium tetrafluoride by reaction with hydrogen fluoride, and the UF/sub 4/ convented to UF/sub 6/ by reaction with a fluorinating agent. The UO/sub 3/ or U/sub 3/O/sub 8/ is placed in a reaction chamber in a copper boat or tray enclosed in a copper oven, and heated to 500 to 650 deg C while hydrogen gas is passed through the oven. The oven is then swept clean of hydrogen and the water vapor formed by means of nitrogen and then while continuing to maintain the temperature between 400 and 600 deg C, anhydrous hydrogen fluoride is passed through. After completion of the conversion to uranium tetrafluoride, the temperature of the reaction chamber is lowered to ahout 400 deg C, and elemental fluorine is used as the fluorinating agent for the conversion of UF/sub 4/ into UF/sub 6/. The fluorine gas is passed into the chamber, and the UF/sub 6/ formed passes out and is delivered to a condenser.

OPTIMIZATION OF HETEROGENEOUS UTILIZATION OF THORIUM IN PWRS TO ENHANCE PROLIFERATION RESISTANCE AND REDUCE WASTE.

DOE Office of Scientific and Technical Information (OSTI.GOV)

TODOSOW,M.; KAZIMI,M.

2004-08-01

Issues affecting the implementation, public perception and acceptance of nuclear power include: proliferation, radioactive waste, safety, and economics. The thorium cycle directly addresses the proliferation and waste issues, but optimization studies of core design and fuel management are needed to ensure that it fits within acceptable safety and economic margins. Typical pressurized water reactors, although loaded with uranium fuel, produce 225 to 275 kg of plutonium per gigawatt-year of operation. Although the spent fuel is highly radioactive, it nevertheless offers a potential proliferation pathway because the plutonium is relatively easy to separate, amounts to many critical masses, and does notmore » present any significant intrinsic barrier to weapon assembly. Uranium 233, on the other hand, produced by the irradiation of thorium, although it too can be used in weapons, may be ''denatured'' by the addition of natural, depleted or low enriched uranium. Furthermore, it appears that the chemical behavior of thoria or thoria-urania fuel makes it a more stable medium for the geological disposal of the spent fuel. It is therefore particularly well suited for a once-through fuel cycle. The use of thorium as a fertile material in nuclear fuel has been of interest since the dawn of nuclear power technology due to its abundance and to potential neutronic advantages. Early projects include homogeneous mixtures of thorium and uranium oxides in the BORAX-IV, Indian Point I, and Elk River reactors, as well as heterogeneous mixtures in the Shippingport seed-blanket reactor. However these projects were developed under considerably different circumstances than those which prevail at present. The earlier applications preceded the current proscription, for non-proliferation purposes, of the use of uranium enriched to more than 20 w/o in {sup 235}U, and has in practice generally prohibited the use of uranium highly enriched in {sup 235}U. They were designed when the expected burnup of light water fuel was on the order of 25 MWD/kgU--about half the present day value--and when it was expected that the spent fuel would be recycled to recover its fissile content.« less

Mineral and energy resources of the Roswell Resource Area, East-Central New Mexico

USGS Publications Warehouse

Bartsch-Winkler, Susan B.; Donatich, Alessandro J.

1995-01-01

The sedimentary formations of the Roswell Resource Area have significant mineral and energy resources. Some of the pre-Pennsylvanian sequences in the Northwestern Shelf of the Permian Basin are oil and gas reservoirs, and Pennsylvanian rocks in Tucumcari Basin are reservoirs of oil and gas as well as source rocks for oil and gas in Triassic rocks. Pre-Permian rocks also contain minor deposits of uranium and vanadium, limestone, and gases. Hydrocarbon reservoirs in Permian rocks include associated gases such as carbon dioxide, helium, and nitrogen. Permian rocks are mineralized adjacent to the Lincoln County porphyry belt, and include deposits of copper, uranium, manganese, iron, polymetallic veins, and Mississippi-Valley-type lead-zinc. Industrial minerals in Permian rocks include fluorite, barite, potash, halite, polyhalite, gypsum, anhydrite, sulfur, limestone, dolomite, brine deposits (iodine and bromine), aggregate (sand), and dimension stone. Doubly terminated quartz crystals, called 'Pecos diamonds' and collected as mineral specimens, occur in Permian rocks along the Pecos River. Mesozoic sedimentary rocks are hosts for copper, uranium, and small quantities of gold-silver-tellurium veins, as well as significant deposits of oil and gas, carbon dioxide, asphalt, coal, and dimension stone. Mesozoic rocks contain limited amounts of limestone, gypsum, petrified wood, and clay. Tertiary rocks host ore deposits commonly associated with intrusive rocks, including platinum-group elements, iron skarns, manganese, uranium and vanadium, molybdenum, polymetallic vein deposits, gold-silver-tellurium veins, and thorium-rare-earth veins. Museum-quality quartz crystals are associated with Tertiary intrusive rocks. Industrial minerals in Tertiary rocks include fluorite, vein- and bedded-barite, caliche, limestone, and aggregate. Tertiary and Quaternary sediments host important placer deposits of gold and titanium, and occurrences of silver and uranium. Important industrial commodities include caliche, limestone and dolomite, and aggregate. Quaternary basalt contains sub-ore-grade uranium, scoria, and clay deposits.

U redox fronts and kaolinisation in basement-hosted unconformity-related U ores of the Athabasca Basin (Canada): late U remobilisation by meteoric fluids

NASA Astrophysics Data System (ADS)

Mercadier, Julien; Cuney, Michel; Cathelineau, Michel; Lacorde, Mathieu

2011-02-01

Proterozoic basement-hosted unconformity-related uranium deposits of the Athabasca Basin (Saskatchewan, Canada) were affected by significant uranium redistribution along oxidation-reduction redox fronts related to cold and late meteoric fluid infiltration. These redox fronts exhibit the same mineralogical and geochemical features as the well-studied uranium roll-front deposits in siliclastic rocks. The primary hydrothermal uranium mineralisation (1.6-1.3 Ga) of basement-hosted deposits is strongly reworked to new disseminated ores comprising three distinctly coloured zones: a white-green zone corresponding to the previous clay-rich alteration halo contemporaneous with hydrothermal ores, a uranium front corresponding to the uranium deposition zone of the redox front (brownish zone, rich in goethite) and a hematite-rich red zone marking the front progression. The three zones directly reflect the mineralogical zonation related to uranium oxides (pitchblende), sulphides, iron minerals (hematite and goethite) and alumino-phosphate-sulphate (APS) minerals. The zoning can be explained by processes of dissolution-precipitation along a redox interface and was produced by the infiltration of cold (<50°C) meteoric fluids to the hydrothermally altered areas. U, Fe, Ca, Pb, S, REE, V, Y, W, Mo and Se were the main mobile elements in this process, and their distribution within the three zones was, for most of them, directly dependent on their redox potential. The elements concentrated in the redox fronts were sourced by the alteration of previously crystallised hydrothermal minerals, such as uranium oxides and light rare earth element (LREE)-rich APS. The uranium oxides from the redox front are characterised by LREE-enriched patterns, which differ from those of unconformity-related ores and clearly demonstrate their distinct conditions of formation. Uranium redox front formation is thought to be linked to fluid circulation episodes initiated during the 400-300 Ma period during uplift and erosion of the Athabasca Basin when it was near the Equator and to have been still active during the last million years. A major kaolinisation event was caused by changes in the fluid circulation regime, reworking the primary uranium redox fronts and causing the redistribution of elements originally concentrated in the uranium-enriched meteoric-related redox fronts.

White Paper – Use of LEU for a Space Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Poston, David Irvin; Mcclure, Patrick Ray

Historically space reactors flown or designed for the U.S. and Russia used Highly Enriched Uranium (HEU) for fuel. HEU almost always produces a small and lighter reactor. Since mass increases launch costs or decreases science payloads, HEU was the natural choice. However in today’s environment, the proliferation of HEU has become a major concern for the U.S. government and hence a policy issue. In addition, launch costs are being reduced as the space community moves toward commercial launch vehicles. HEU also carries a heavy security cost to process, test, transport and launch. Together these issues have called for a re-investigationmore » into space reactors the use Low Enriched Uranium (LEU) fuel.« less

PROCESS OF DISSOLVING FUEL ELEMENTS OF NUCLEAR REACTORS

DOEpatents

Wall, E.M.V.; Bauer, D.T.; Hahn, H.T.

1963-09-01

A process is described for dissolving stainless-steelor zirconium-clad uranium dioxide fuel elements by immersing the elements in molten lead chloride, adding copper, cuprous chloride, or cupric chloride as a catalyst and passing chlorine through the salt mixture. (AEC)

A STUDY OF THE ACIDOSIS, BLOOD UREA, AND PLASMA CHLORIDES IN URANIUM NEPHRITIS IN THE DOG, AND OF THE PROTECTIVE ACTION OF SODIUM BICARBONATE.

PubMed

Goto, K

1917-05-01

1. The presence of an acidosis in dogs with experimental uranium nephritis is demonstrable by the Van Slyke-Stillman-Cullen method and that of Marriott. It is detected more readily by the former method. 2. This acidosis is associated with increase in the blood urea and plasma chlorides and with the appearance of albumin and casts in the urine. 3. The oral administration of sodium bicarbonate diminishes the acidosis, the increase in plasma chlorides, the amount of albumin and casts in the urine, and, to a lesser degree, the increase in the blood urea following the administration of uranium. It also diminishes the severity of the changes produced by uranium in the kidneys. 4. The oral administration of sodium bicarbonate to normal dogs raises the carbon dioxide content of the plasma as determined by the. Van Slyke-Stillman-Cullen method.

An unattended verification station for UF 6 cylinders: Field trial findings

DOE PAGES

Smith, L. E.; Miller, K. A.; McDonald, B. S.; ...

2017-08-26

In recent years, the International Atomic Energy Agency (IAEA) has pursued innovative techniques and an integrated suite of safeguards measures to address the verification challenges posed by the front end of the nuclear fuel cycle. Among the unattended instruments currently being explored by the IAEA is an Unattended Cylinder Verification Station (UCVS), which could provide automated, independent verification of the declared relative enrichment, 235U mass, total uranium mass, and identification for all declared uranium hexafluoride cylinders in a facility (e.g., uranium enrichment plants and fuel fabrication plants). Under the auspices of the United States and European Commission Support Programs tomore » the IAEA, a project was undertaken to assess the technical and practical viability of the UCVS concept. The first phase of the UCVS viability study was centered on a long-term field trial of a prototype UCVS system at a fuel fabrication facility. A key outcome of the study was a quantitative performance evaluation of two nondestructive assay (NDA) methods being considered for inclusion in a UCVS: Hybrid Enrichment Verification Array (HEVA), and Passive Neutron Enrichment Meter (PNEM). This paper provides a description of the UCVS prototype design and an overview of the long-term field trial. In conclusion, analysis results and interpretation are presented with a focus on the performance of PNEM and HEVA for the assay of over 200 “typical” Type 30B cylinders, and the viability of an “NDA Fingerprint” concept as a high-fidelity means to periodically verify that material diversion has not occurred.« less

An unattended verification station for UF 6 cylinders: Field trial findings

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, L. E.; Miller, K. A.; McDonald, B. S.

In recent years, the International Atomic Energy Agency (IAEA) has pursued innovative techniques and an integrated suite of safeguards measures to address the verification challenges posed by the front end of the nuclear fuel cycle. Among the unattended instruments currently being explored by the IAEA is an Unattended Cylinder Verification Station (UCVS), which could provide automated, independent verification of the declared relative enrichment, 235U mass, total uranium mass, and identification for all declared uranium hexafluoride cylinders in a facility (e.g., uranium enrichment plants and fuel fabrication plants). Under the auspices of the United States and European Commission Support Programs tomore » the IAEA, a project was undertaken to assess the technical and practical viability of the UCVS concept. The first phase of the UCVS viability study was centered on a long-term field trial of a prototype UCVS system at a fuel fabrication facility. A key outcome of the study was a quantitative performance evaluation of two nondestructive assay (NDA) methods being considered for inclusion in a UCVS: Hybrid Enrichment Verification Array (HEVA), and Passive Neutron Enrichment Meter (PNEM). This paper provides a description of the UCVS prototype design and an overview of the long-term field trial. In conclusion, analysis results and interpretation are presented with a focus on the performance of PNEM and HEVA for the assay of over 200 “typical” Type 30B cylinders, and the viability of an “NDA Fingerprint” concept as a high-fidelity means to periodically verify that material diversion has not occurred.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Arnason, John G.

Background: Between 1958 and 1982, NL Industries manufactured components of enriched (EU) and depleted uranium (DU) at a factory in Colonie NY, USA. More than 5 metric tons of DU was deposited as microscopic DU oxide particles on the plant site and surrounding residential community. A prior study involving a small number of individuals (n=23) indicated some residents were exposed to DU and former workers to both DU and EU, most probably through inhalation of aerosol particles. Objectives: Our aim was to measure total uranium [U] and the uranium isotope ratios: {sup 234}U/{sup 238}U; {sup 235}U/{sup 238}U; and {sup 236}U/{supmore » 238}U, in the urine of a cohort of former workers and nearby residents of the NLI factory, to characterize individual exposure to natural uranium (NU), DU, and EU more than 3 decades after production ceased. Methods: We conducted a biomonitoring study in a larger cohort of 32 former workers and 99 residents, who may have been exposed during its period of operation, by measuring Total U, NU, DU, and EU in urine using Sector Field Inductively Coupled Plasma - Mass Spectrometry (SF-ICP-MS). Results: Among workers, 84% were exposed to DU, 9% to EU and DU, and 6% to natural uranium (NU) only. For those exposed to DU, urinary isotopic and [U] compositions result from binary mixing of NU and the DU plant feedstock. Among residents, 8% show evidence of DU exposure, whereas none shows evidence of EU exposure. For residents, the [U] geometric mean is significantly below the value reported for NHANES. There is no significant difference in [U] between exposed and unexposed residents, suggesting that [U] alone is not a reliable indicator of exposure to DU in this group. Conclusions: Ninety four percent of workers tested showed evidence of exposure to DU, EU or both, and were still excreting DU and EU decades after leaving the workforce. The study demonstrates the advantage of measuring multiple isotopic ratios (e.g., {sup 236}U/{sup 238}U and {sup 235}U/{sup 238}U) over a single ratio ({sup 235}U/{sup 238}U) in determining sources of uranium exposure. - Highlights: • Biomonitoring study of residents and former workers exposed to DU in Colonie NY. • Urine (99 residents+32 former workers) analyzed for depleted uranium (DU). • DU detected in 84% of workers and 8% of residents >30 years after plant closed. • Enriched uranium detected in 6% of former workers based on isotope ratios.« less

Organic geochemical analysis of sedimentary organic matter associated with uranium

USGS Publications Warehouse

Leventhal, J.S.; Daws, T.A.; Frye, J.S.

1986-01-01

Samples of sedimentary organic matter from several geologic environments and ages which are enriched in uranium (56 ppm to 12%) have been characterized. The three analytical techniqyes used to study the samples were Rock-Eval pyrolysis, pyrolysis-gas chromatography-mass spectrometry, and solid-state C-13 nuclear magnetic resonance (NMR) spectroscopy. In samples with low uranium content, the pyrolysis-gas chromatography products contain oxygenated functional groups (as hydroxyl) and molecules with both aliphatic and aromatic carbon atoms. These samples with low uranium content give measurable Rock-Eval hydrocarbon and organic-CO2 yields, and C-13 NMR values of > 30% aliphatic carbon. In contrast, uranium-rich samples have few hydrocarbon pyrolysis products, increased Rock-Eval organic-CO2 contents and > 70% aromatic carbon contents from C-13 NMR. The increase in aromaticity and decrease in hydrocarbon pyrolysis yield are related to the amount of uranium and the age of the uranium minerals, which correspond to the degree of radiation damage. The three analytical techniques give complementary results. Increase in Rock-Eval organic-CO2 yield correlates with uranium content for samples from the Grants uranium region. Calculations show that the amount of organic-CO2 corresponds to the quantity of uranium chemically reduced by the organic matter for the Grants uranium region samples. ?? 1986.

Sintering Uranium Dioxide of Domestic Production. Report No. 78; SINTERIZACION DE DIOXIDO DE URANIO DE PRODUCCION NACIONAL. Informe No. 78

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carrea, A.J.

1963-01-01

After a brief indication of the uranium- oxygen equilibrium and the methods for the preparation of UO/sub 2/, the sintering of UO/sub 2/ is considered. The effects of various sintering atmospheres on the properties of the product are discussed and tabulated. The method used for the processing of domestic ores for the preparation of UO/sub 2/ and the fabricition of the sintered UO/sub 2/are described. The properties of the product obtained are illustrated graphically. (J.S.R.)

Thermodynamic calculations of oxygen self-diffusion in mixed-oxide nuclear fuels

DOE PAGES

Parfitt, David C.; Cooper, Michael William; Rushton, Michael J.D.; ...

2016-07-29

Mixed-oxide fuels containing uranium with thorium and/or plutonium may play an important part in future nuclear fuel cycles. There are, however, significantly less data available for these materials than conventional uranium dioxide fuel. In the present study, we employ molecular dynamics calculations to simulate the elastic properties and thermal expansivity of a range of mixed oxide compositions. These are then used to support equations of state and oxygen self-diffusion models to provide a self-consistent prediction of the behaviour of these mixed oxide fuels at arbitrary compositions.

Infrared spectroscopy of extreme coordination: the carbonyls of U(+) and UO(2)(+).

PubMed

Ricks, Allen M; Gagliardi, Laura; Duncan, Michael A

2010-11-17

Uranium and uranium dioxide carbonyl cations produced by laser vaporization are studied with mass-selected ion infrared spectroscopy in the C-O stretching region. Dissociation patterns, spectra, and quantum chemical calculations establish that the fully coordinated ions are U(CO)(8)(+) and UO(2)(CO)(5)(+), with D(4d) square antiprism and D(5h) pentagonal bipyramid structures. Back-bonding in U(CO)(8)(+) causes a red-shifted CO stretch, but back-donation is inefficient for UO(2)(CO)(5)(+), producing a blue-shifted CO stretch characteristic of nonclassical carbonyls.

Occupational safety data and casualty rates for the uranium fuel cycle. [Glossaries

DOE Office of Scientific and Technical Information (OSTI.GOV)

O'Donnell, F.R.; Hoy, H.C.

1981-10-01

Occupational casualty (injuries, illnesses, fatalities, and lost workdays) and production data are presented and used to calculate occupational casualty incidence rates for technologies that make up the uranium fuel cycle, including: mining, milling, conversion, and enrichment of uranium; fabrication of reactor fuel; transportation of uranium and fuel elements; generation of electric power; and transmission of electric power. Each technology is treated in a separate chapter. All data sources are referenced. All steps used to calculate normalized occupational casualty incidence rates from the data are presented. Rates given include fatalities, serious cases, and lost workdays per 100 man-years worked, per 10/supmore » 12/ Btu of energy output, and per other appropriate units of output.« less

Special nuclear material simulation device

DOEpatents

Leckey, John H.; DeMint, Amy; Gooch, Jack; Hawk, Todd; Pickett, Chris A.; Blessinger, Chris; York, Robbie L.

2014-08-12

An apparatus for simulating special nuclear material is provided. The apparatus typically contains a small quantity of special nuclear material (SNM) in a configuration that simulates a much larger quantity of SNM. Generally the apparatus includes a spherical shell that is formed from an alloy containing a small quantity of highly enriched uranium. Also typically provided is a core of depleted uranium. A spacer, typically aluminum, may be used to separate the depleted uranium from the shell of uranium alloy. A cladding, typically made of titanium, is provided to seal the source. Methods are provided to simulate SNM for testing radiation monitoring portals. Typically the methods use at least one primary SNM spectral line and exclude at least one secondary SNM spectral line.

Fuel preparation for use in the production of medical isotopes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Policke, Timothy A.; Aase, Scott B.; Stagg, William R.

The present invention relates generally to the field of medical isotope production by fission of uranium-235 and the fuel utilized therein (e.g., the production of suitable Low Enriched Uranium (LEU is uranium having 20 weight percent or less uranium-235) fuel for medical isotope production) and, in particular to a method for producing LEU fuel and a LEU fuel product that is suitable for use in the production of medical isotopes. In one embodiment, the LEU fuel of the present invention is designed to be utilized in an Aqueous Homogeneous Reactor (AHR) for the production of various medical isotopes including, butmore » not limited to, molybdenum-99, cesium-137, iodine-131, strontium-89, xenon-133 and yttrium-90.« less

Irans Nuclear Program: Tehrans Compliance with International Obligations

DTIC Science & Technology

2016-04-07

ratified the nuclear Nonproliferation Treaty (NPT) in 1970. Article III of the treaty requires non-nuclear- weapon states-parties 1 to accept...concern that Tehran is pursuing nuclear weapons . Tehran’s construction of gas centrifuge uranium enrichment facilities is currently the main source...uranium (HEU), which is one of the two types of fissile material used in nuclear weapons . HEU can also be used as fuel in certain types of nuclear

Irans Nuclear Program: Tehrans Compliance with International Obligations

DTIC Science & Technology

2016-03-03

ratified the nuclear Nonproliferation Treaty (NPT) in 1970. Article III of the treaty requires non-nuclear- weapon states-parties 1 to accept...concern that Tehran is pursuing nuclear weapons . Tehran’s construction of gas centrifuge uranium enrichment facilities is currently the main source...uranium (HEU), which is one of the two types of fissile material used in nuclear weapons . HEU can also be used as fuel in certain types of nuclear

JPRS Report, Science & Technology, Japan

DTIC Science & Technology

1987-11-12

Change (4) Future Direction Anyway, it has become almost clear that the effect of power recovery cannot be expected from the insulation of...process spent fuels in greater safety and to recover the uranium or plutonium from spent fuels for effective reapplication. In 1974, the PNC began...constructed to serve as a pilot plant that could be used to establish reprocessing technology for the next practical stage. 32 As for enriched uranium

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bakel, Allen J.; Conner, Cliff; Quigley, Kevin

One of the missions of the Reduced Enrichment for Research and Test Reactors (RERTR) program (and now the National Nuclear Security Administrations Material Management and Minimization program) is to facilitate the use of low enriched uranium (LEU) targets for 99Mo production. The conversion from highly enriched uranium (HEU) to LEU targets will require five to six times more uranium to produce an equivalent amount of 99Mo. The work discussed here addresses the technical challenges encountered in the treatment of uranyl nitrate hexahydrate (UNH)/nitric acid solutions remaining after the dissolution of LEU targets. Specifically, the focus of this work is themore » calcination of the uranium waste from 99Mo production using LEU foil targets and the Modified Cintichem Process. Work with our calciner system showed that high furnace temperature, a large vent tube, and a mechanical shield are beneficial for calciner operation. One- and two-step direct calcination processes were evaluated. The high-temperature one-step process led to contamination of the calciner system. The two-step direct calcination process operated stably and resulted in a relatively large amount of material in the calciner cup. Chemically assisted calcination using peroxide was rejected for further work due to the difficulty in handling the products. Chemically assisted calcination using formic acid was rejected due to unstable operation. Chemically assisted calcination using oxalic acid was recommended, although a better understanding of its chemistry is needed. Overall, this work showed that the two-step direct calcination and the in-cup oxalic acid processes are the best approaches for the treatment of the UNH/nitric acid waste solutions remaining from dissolution of LEU targets for 99Mo production.« less

Assessment of Nuclear Fuels using Radiographic Thickness Measurement Method

DOE Office of Scientific and Technical Information (OSTI.GOV)

Muhammad Abir; Fahima Islam; Hyoung Koo Lee

2014-11-01

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

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

1997-08-01

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

10 CFR Appendix E to Part 110 - Illustrative List of Chemical Exchange or Ion Exchange Enrichment Plant Equipment and Components...

Code of Federal Regulations, 2012 CFR

2012-01-01

.... A major design concern is to avoid contamination of the process streams with certain metal ions... internal turbine mixers), especially designed or prepared for uranium enrichment using the chemical...) or glass. The stage residence time of the columns is designed to be short (30 seconds or less). (2...

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

Federal Register 2010, 2011, 2012, 2013, 2014

2012-03-27

... that cascades number 1.5, 1.6, 1.7, 1.8, 2.1, and 2.4 as well as autoclave one of the facility have... 2.4 as well as autoclave one of the facility have been constructed in accordance with the... Facility Inspection Reports Regarding Louisiana Energy Services LLC, National Enrichment Facility, Eunice...

10 CFR Appendix B to Part 110 - Illustrative List of Gas Centrifuge Enrichment Plant Components Under NRC's Export Licensing...

Code of Federal Regulations, 2010 CFR

2010-01-01

... materials of construction for the rotating rotor assembly, and hence its individual components, have to be... gas centrifuge for uranium enrichment is characterized by having within the rotor chamber a rotating... featuring at least 3 separate channels of which 2 are connected to scoops extending from the rotor axis...

10 CFR 70.60 - Applicability.

Code of Federal Regulations, 2011 CFR

2011-01-01

... Certain Licensees Authorized To Possess a Critical Mass of Special Nuclear Material § 70.60 Applicability... critical mass of special nuclear material, and engaged in enriched uranium processing, fabrication of...

Identifying anthropogenic uranium compounds using soft X-ray near-edge absorption spectroscopy

NASA Astrophysics Data System (ADS)

Ward, Jesse D.; Bowden, Mark; Tom Resch, C.; Eiden, Gregory C.; Pemmaraju, C. D.; Prendergast, David; Duffin, Andrew M.

2017-01-01

Uranium ores mined for industrial use are typically acid-leached to produce yellowcake and then converted into uranium halides for enrichment and purification. These anthropogenic chemical forms of uranium are distinct from their mineral counterparts. The purpose of this study is to use soft X-ray absorption spectroscopy to characterize several common anthropogenic uranium compounds important to the nuclear fuel cycle. Chemical analyses of these compounds are important for process and environmental monitoring. X-ray absorption techniques have several advantages in this regard, including element-specificity, chemical sensitivity, and high spectral resolution. Oxygen K-edge spectra were collected for uranyl nitrate, uranyl fluoride, and uranyl chloride, and fluorine K-edge spectra were collected for uranyl fluoride and uranium tetrafluoride. Interpretation of the data is aided by comparisons to calculated spectra. The effect of hydration state on the sample, a potential complication in interpreting oxygen K-edge spectra, is discussed. These compounds have unique spectral signatures that can be used to identify unknown samples.

Column bioleaching of uranium embedded in granite porphyry by a mesophilic acidophilic consortium.

PubMed

Qiu, Guanzhou; Li, Qian; Yu, Runlan; Sun, Zhanxue; Liu, Yajie; Chen, Miao; Yin, Huaqun; Zhang, Yage; Liang, Yili; Xu, Lingling; Sun, Limin; Liu, Xueduan

2011-04-01

A mesophilic acidophilic consortium was enriched from acid mine drainage samples collected from several uranium mines in China. The performance of the consortium in column bioleaching of low-grade uranium embedded in granite porphyry was investigated. The influences of several chemical parameters on uranium extraction in column reactor were also investigated. A uranium recovery of 96.82% was achieved in 97 days column leaching process including 33 days acid pre-leaching stage and 64 days bioleaching stage. It was reflected that indirect leaching mechanism took precedence over direct. Furthermore, the bacterial community structure was analyzed by using Amplified Ribosomal DNA Restriction Analysis. The results showed that microorganisms on the residual surface were more diverse than that in the solution. Acidithiobacillus ferrooxidans was the dominant species in the solution and Leptospirillum ferriphilum on the residual surface. Copyright © 2011 Elsevier Ltd. All rights reserved.

METHOD OF FABRICATING A URANIUM-ZIRCONIUM HYDRIDE REACTOR CORE

DOEpatents

Weeks, I.F.; Goeddel, W.V.

1960-03-22

A method is described of evenly dispersing uranlum metal in a zirconium hydride moderator to produce a fuel element for nuclear reactors. According to the invention enriched uranium hydride and zirconium hydride powders of 200 mesh particle size are thoroughly admixed to form a mixture containing 0.1 to 3% by weight of U/sup 235/ hydride. The mixed powders are placed in a die and pressed at 100 tons per square inch at room temperature. The resultant compacts are heated in a vacuum to 300 deg C, whereby the uranium hydride deoomposes into uranium metal and hydrogen gas. The escaping hydrogen gas forms a porous matrix of zirconium hydride, with uramum metal evenly dispersed therethrough. The advantage of the invention is that the porosity and uranium distribution of the final fuel element can be more closely determined and controlled than was possible using prior methods of producing such fuel ele- ments.

Physics and potentials of fissioning plasmas for space power and propulsion

NASA Technical Reports Server (NTRS)

Thom, K.; Schwenk, F. C.; Schneider, R. T.

1976-01-01

Fissioning uranium plasmas are the nuclear fuel in conceptual high-temperature gaseous-core reactors for advanced rocket propulsion in space. A gaseous-core nuclear rocket would be a thermal reactor in which an enriched uranium plasma at about 10,000 K is confined in a reflector-moderator cavity where it is nuclear critical and transfers its fission power to a confining propellant flow for the production of thrust at a specific impulse up to 5000 sec. With a thrust-to-engine weight ratio approaching unity, the gaseous-core nuclear rocket could provide for propulsion capabilities needed for manned missions to the nearby planets and for economical cislunar ferry services. Fueled with enriched uranium hexafluoride and operated at temperatures lower than needed for propulsion, the gaseous-core reactor scheme also offers significant benefits in applications for space and terrestrial power. They include high-efficiency power generation at low specific mass, the burnup of certain fission products and actinides, the breeding of U-233 from thorium with short doubling times, and improved convenience of fuel handling and processing in the gaseous phase.

Measurement of the 19F(α,n)22Na Cross Section for Nuclear Safeguards Science

NASA Astrophysics Data System (ADS)

Lowe, Marcus; Smith, M. S.; Pain, S.; Febbraro, M.; Pittman, S.; Chipps, K. A.; Thompson, S. J.; Grinder, M.; Grzywacz, R.; Smith, K.; Thornsberry, C.; Thompson, P.; Peters, W. A.; Waddell, D.; Blanchard, R.; Carls, A.; Shadrick, S.; Engelhardt, A.; Hertz-Kintish, D.; Allen, N.; Sims, H.

2015-10-01

Enriched uranium is commonly stored in fluoride matrices such as UF6. Alpha decays of uranium in UF6 will create neutrons via the 19F(α,n)22Na reaction. An improved cross section for this reaction will enable improved nondestructive assays of uranium content in storage cylinders at material enrichment facilities. To determine this reaction cross section, we have performed experiments using both forward and inverse kinematic techniques at the University of Notre Dame (forward) and Oak Ridge National Laboratory (inverse). Both experiments utilized the Versatile Array of Neutron Detectors at Low Energy (VANDLE) for neutron detection. The ORNL experiment also used a new ionization chamber for 22Na particle identification. Gating on the 22Na nuclei detected drastically reduced the background counts in the neutron time-of-flight spectra. The latest analysis and results will be presented for 19F beam energies ranging from 20-37 MeV. This work is funded in part by the DOE Office of Nuclear Physics, the National Nuclear Security Administration's Office of Defense Nuclear Nonproliferation R&D, and the NSF.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fishbone, L.G.; Moussalli, G.; Naegele, G.

1994-04-01

An approach of short-notice random inspections (SNRIs) for inventory-change verification can enhance the effectiveness and efficiency of international safeguards at natural or low-enriched uranium (LEU) fuel fabrication plants. According to this approach, the plant operator declares the contents of nuclear material items before knowing if an inspection will occur to verify them. Additionally, items about which declarations are newly made should remain available for verification for an agreed time. This report details a six-month field test of the feasibility of such SNRIs which took place at the Westinghouse Electric Corporation Commercial Nuclear Fuel Division. Westinghouse personnel made daily declarations aboutmore » both feed and product items, uranium hexafluoride cylinders and finished fuel assemblies, using a custom-designed computer ``mailbox``. Safeguards inspectors from the IAEA conducted eight SNRIs to verify these declarations. Items from both strata were verified during the SNRIs by means of nondestructive assay equipment. The field test demonstrated the feasibility and practicality of key elements of the SNRI approach for a large LEU fuel fabrication plant.« less

A cellular automaton method to simulate the microstructure and evolution of low-enriched uranium (LEU) U–Mo/Al dispersion type fuel plates

DOE Office of Scientific and Technical Information (OSTI.GOV)

Drera, Saleem S.; Hofman, Gerard L.; Kee, Robert J.

Low-enriched uranium (LEU) fuel plates for high power materials test reactors (MTR) are composed of nominally spherical uranium-molybdenum (U-Mo) particles within an aluminum matrix. Fresh U-Mo particles typically range between 10 and 100 mu m in diameter, with particle volume fractions up to 50%. As the fuel ages, reaction-diffusion processes cause the formation and growth of interaction layers that surround the fuel particles. The growth rate depends upon the temperature and radiation environment. The cellular automaton algorithm described in this paper can synthesize realistic random fuel-particle structures and simulate the growth of the intermetallic interaction layers. Examples in the presentmore » paper pack approximately 1000 particles into three-dimensional rectangular fuel structures that are approximately 1 mm on each side. The computational approach is designed to yield synthetic microstructures consistent with images from actual fuel plates and is validated by comparison with empirical data on actual fuel plates. (C) 2014 Elsevier B.V. All rights reserved.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kelly Cummins; Igor Bolshinsky; Ken Allen

2009-07-01

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

Transient analysis for the tajoura critical facility with IRT-2M HEU fuel and IRT-4M leu fuel : ANL independent verification results.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Garner, P. L.; Hanan, N. A.

2005-12-02

Calculations have been performed for postulated transients in the Critical Facility at the Tajoura Nuclear Research Center (TNRC) in Libya. These calculations have been performed at the request of staff of the Renewable Energy and Water Desalinization Research Center (REWDRC) who are performing similar calculations. The transients considered were established during a working meeting between ANL and REWDRC staff on October 1-2, 2005 and subsequent email correspondence. Calculations were performed for the current high-enriched uranium (HEU) core and the proposed low-enriched uranium (LEU) core. These calculations have been performed independently from those being performed by REWDRC and serve as onemore » step in the verification process.« less

Investigating microstructural evolution during the electroreduction of UO2 to U in LiCl-KCl eutectic using focused ion beam tomography

NASA Astrophysics Data System (ADS)

Brown, L. D.; Abdulaziz, R.; Tjaden, B.; Inman, D.; Brett, D. J. L.; Shearing, P. R.

2016-11-01

Reprocessing of spent nuclear fuels using molten salt media is an attractive alternative to liquid-liquid extraction techniques. Pyroelectrochemical processing utilizes direct, selective, electrochemical reduction of uranium dioxide, followed by selective electroplating of a uranium metal. Thermodynamic prediction of the electrochemical reduction of UO2 to U in LiCl-KCl eutectic has shown to be a function of the oxide ion activity. The pO2- of the salt may be affected by the microstructure of the UO2 electrode. A uranium dioxide filled "micro-bucket" electrode has been partially electroreduced to uranium metal in molten lithium chloride-potassium chloride eutectic. This partial electroreduction resulted in two distinct microstructures: a dense UO2 and a porous U metal structure were characterised by energy dispersive X-ray spectroscopy. Focused ion beam tomography was performed on five regions of this electrode which revealed an overall porosity ranging from 17.36% at the outer edge to 3.91% towards the centre, commensurate with the expected extent of reaction in each location. The pore connectivity was also seen to reduce from 88.32% to 17.86% in the same regions and the tortuosity through the sample was modelled along the axis of propagation of the electroreduction, which was seen to increase from a value of 4.42 to a value of infinity (disconnected pores). These microstructural characteristics could impede the transport of O2- ions resulting in a change in the local pO2- which could result in the inability to perform the electroreduction.