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...

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.

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...

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...

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...

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...

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...

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...

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...

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

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

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...

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...

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...

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...

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

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...

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

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.

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...

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...

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.

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...

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

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...

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

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)

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.

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

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

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...

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

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

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...

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

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

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

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

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...

U-235 Holdup Measurements in the 321-M Lathe HEPA Banks

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 Decommissioning 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 results of the holdup assays are essential for determining compliance with the Waste Acceptance Criteria, Material Control and Accountability, and to meet criticality safety controls. This report covers holdup measurements of uranium residue in six high efficiency particulate air (HEPA) filter banks of the A-lathe and B-lathe exhaust systems of the 321-M facility. This report discusses the non-destructive assay measurements,more » assumptions, calculations, and results of the uranium holdup in these six items.« less

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

China and Proliferation of Weapons of Mass Destruction and Missiles: Policy Issues

DTIC Science & Technology

2012-11-07

facilities, while experts from China worked at a uranium mine at Saghand and a centrifuge facility (for uranium enrichment) near Isfahan, reported the...Barbara Opall -Rome and Vago Muradian, “Bush Privately Lauds Israeli Attack on Syria,” Defense News, January 14, 2008; Paul Richter, “West Says N... Mining Development Trading Corporation).123 Also, in December 2009, Japan arrested two traders who exported expensive cosmetics from Japan to North

China and Proliferation of Weapons of Mass Destruction and Missiles: Policy Issues

DTIC Science & Technology

2012-03-30

from China worked at a uranium mine at Saghand and a centrifuge facility (for uranium enrichment) near Isfahan, reported the Washington Post (December...Facilities,” China News Agency, September 3, 2007; Xinhua, September 4 and 6, 2007. 99 Barbara Opall -Rome and Vago Muradian, “Bush Privately Lauds...with the DPRK’s arms dealer, Global Trading and Technology (a front for Korea Mining Development Trading Corporation).119 Also, in December 2009

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...

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

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

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

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

10 CFR Appendix D to Subpart D of... - Classes of Actions That Normally Require EISs

Code of Federal Regulations, 2010 CFR

2010-01-01

... average megawatts or more over a 12 month period. This applies to power marketing operations and to siting... Systems D2. Siting/construction/operation/decommissioning of nuclear fuel reprocessing facilities D3. Siting/construction/operation/decommissioning of uranium enrichment facilities D4. Siting/construction...

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

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.

Paducah Site annual report for 1995

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

Belcher, G.

1997-01-01

The Paducah Gaseous Diffusion Plant, located in McCracken County, Kentucky, has been producing enriched uranium since 1952. In July 1993, the US department of Energy (DOE) leased the production areas of the site to the US Enrichment Corporation (USEC). A new subsidiary of Lockheed Martin Corporation, Lockheed Martin Utility Services, manages the leased facilities for USEC. DOE maintains responsibility for the environmental restoration, waste management, and enrichment facilities activities at the plant through its management contractor, Lockheed Martin Energy Systems. The purpose of this document is to summarize calendar year 1995 environmental monitoring activities for DOE activities at the Paducahmore » Site. DOE requires all of its facilities to conduct and document such activities annually. This report does not include USEC environmental activities.« 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

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.

Flowsheets and source terms for radioactive waste projections

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

Forsberg, C.W.

1985-03-01

Flowsheets and source terms used to generate radioactive waste projections in the Integrated Data Base (IDB) Program are given. Volumes of each waste type generated per unit product throughput have been determined for the following facilities: uranium mining, UF/sub 6/ conversion, uranium enrichment, fuel fabrication, boiling-water reactors (BWRs), pressurized-water reactors (PWRs), and fuel reprocessing. Source terms for DOE/defense wastes have been developed. Expected wastes from typical decommissioning operations for each facility type have been determined. All wastes are also characterized by isotopic composition at time of generation and by general chemical composition. 70 references, 21 figures, 53 tables.

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

31 CFR 540.316 - Uranium enrichment.

Code of Federal Regulations, 2013 CFR

2013-07-01

... 31 Money and Finance:Treasury 3 2013-07-01 2013-07-01 false Uranium enrichment. 540.316 Section... FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.316 Uranium enrichment. The term uranium enrichment means the process of...

31 CFR 540.316 - Uranium enrichment.

Code of Federal Regulations, 2014 CFR

2014-07-01

... 31 Money and Finance:Treasury 3 2014-07-01 2014-07-01 false Uranium enrichment. 540.316 Section... FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.316 Uranium enrichment. The term uranium enrichment means the process of...

31 CFR 540.316 - Uranium enrichment.

Code of Federal Regulations, 2011 CFR

2011-07-01

... 31 Money and Finance:Treasury 3 2011-07-01 2011-07-01 false Uranium enrichment. 540.316 Section... FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.316 Uranium enrichment. The term uranium enrichment means the process of...

31 CFR 540.316 - Uranium enrichment.

Code of Federal Regulations, 2012 CFR

2012-07-01

... 31 Money and Finance:Treasury 3 2012-07-01 2012-07-01 false Uranium enrichment. 540.316 Section... FOREIGN ASSETS CONTROL, DEPARTMENT OF THE TREASURY HIGHLY ENRICHED URANIUM (HEU) AGREEMENT ASSETS CONTROL REGULATIONS General Definitions § 540.316 Uranium enrichment. The term uranium enrichment means the process of...

Secretary Chu and Administrator D'Agostino at the HEUMF Dedication

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

Department of Energy Secretary Steven Chu and Administrator D'Agostino

2010-03-24

Speeches by Secretary of Energy Chu and NNSA Administrator Tom D'Agostino at the Highly Enriched Uranium Materials Facility dedication ceremony at the Y-12 National Security Complex in Oak Ridge, Tennessee on March 22, 2010.

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

16. VIEW OF THE ENRICHED URANIUM RECOVERY SYSTEM. ENRICHED URANIUM ...

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

16. VIEW OF THE ENRICHED URANIUM RECOVERY SYSTEM. ENRICHED URANIUM RECOVERY PROCESSED RELATIVELY PURE MATERIALS AND SOLUTIONS AND SOLID RESIDUES WITH RELATIVELY LOW URANIUM CONTENT. URANIUM RECOVERY INVOLVED BOTH SLOW AND FAST PROCESSES. (4/4/66) - Rocky Flats Plant, General Manufacturing, Support, Records-Central Computing, Southern portion of Plant, Golden, Jefferson County, CO

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

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

31 CFR 540.306 - Highly Enriched Uranium (HEU).

Code of Federal Regulations, 2011 CFR

2011-07-01

... 31 Money and Finance:Treasury 3 2011-07-01 2011-07-01 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 540.306 - 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...

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

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...

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.

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...

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

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

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...

Feasibility of a Fieldable Mass Spectrometer FY 2015 Year-end Report

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

Barinaga, Charles J.; Hager, George J.; Hoegg, Edward D.

Currently, the International Atomic Energy Agency (IAEA) monitors the production of enriched uranium hexafluoride (UF 6) at declared facilities by collecting a few grams of product in sample tubes that are then sent to central laboratories for processing and isotope ratio analysis by thermal ionization mass spectrometry. Analysis of results may not be available for some time after collection. In addition, new shipping regulations will make it more difficult to transport this amount of UF 6 to a laboratory. The IAEA is interested in an isotope ratio technique for uranium in UF 6 that can be moved to and operatedmore » at the enrichment facility itself. This report covers the tasks and activities of the Feasibility of a Fieldable Mass Spectrometer Project for FY 2015, which investigates the feasibility of an in-field isotope ratio technique— the forward deployment of a technique to the non-laboratory situation of a protected room with power and heat at the facility of interest. A variety of nontraditional elemental ionization techniques were considered. It was determined that only two of these should be moved forward for testing with the candidate in-field mass spectrometer and with the adsorbed UF 6 sample types.« 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

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.

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

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

Validation of gamma-ray detection techniques for safeguards monitoring at natural uranium conversion facilities

NASA Astrophysics Data System (ADS)

Dewji, S. A.; Lee, D. L.; Croft, S.; Hertel, N. E.; Chapman, J. A.; McElroy, R. D.; Cleveland, S.

2016-07-01

Recent IAEA circulars and policy papers have sought to implement safeguards when any purified aqueous uranium solution or uranium oxides suitable for isotopic enrichment or fuel fabrication exists. Under the revised policy, IAEA Policy Paper 18, the starting point for nuclear material under safeguards was reinterpreted, suggesting that purified uranium compounds should be subject to safeguards procedures no later than the first point in the conversion process. In response to this technical need, a combination of simulation models and experimental measurements were employed to develop and validate concepts of nondestructive assay monitoring systems in a natural uranium conversion plant (NUCP). In particular, uranyl nitrate (UO2(NO3)2) solution exiting solvent extraction was identified as a key measurement point (KMP), where gamma-ray spectroscopy was selected as the process monitoring tool. The Uranyl Nitrate Calibration Loop Equipment (UNCLE) facility at Oak Ridge National Laboratory was employed to simulate the full-scale operating conditions of a purified uranium-bearing aqueous stream exiting the solvent extraction process in an NUCP. Nondestructive assay techniques using gamma-ray spectroscopy were evaluated to determine their viability as a technical means for drawing safeguards conclusions at NUCPs, and if the IAEA detection requirements of 1 significant quantity (SQ) can be met in a timely way. This work investigated gamma-ray signatures of uranyl nitrate circulating in the UNCLE facility and evaluated various gamma-ray detector sensitivities to uranyl nitrate. These detector validation activities include assessing detector responses to the uranyl nitrate gamma-ray signatures for spectrometers based on sodium iodide, lanthanum bromide, and high-purity germanium detectors. The results of measurements under static and dynamic operating conditions at concentrations ranging from 10-90 g U/L of natural uranyl nitrate are presented. A range of gamma-ray lines is examined, including attenuation for transmission measurement of density and concentration. It was determined that transmission-corrected gamma-ray spectra provide a reliable way to monitor the 235U concentration of uranyl nitrate solution in transfer pipes in NUCPs. Furthermore, existing predictive and analysis methods are adequate to design and realize practical designs. The 137Cs transmission source employed in this work is viable but not optimal for 235U densitometry determination. Validated simulations assessed the viability of 133Ba and 57Co as alternative densitometry sources. All three gamma-ray detectors are viable for monitoring natural uranium feed; although high-purity germanium is easiest to interpret, it is, however, the least attractive as an installation instrument. Overall, for monitoring throughput in a facility such as UNCLE, emulating the uranium concentration and pump speeds of the Springfields conversion facility in the United Kingdom, an uncertainty of less than 0.17% is required in order to detect the diversion of 1 SQ of uranyl nitrate through changes in uranium concentration over an accountancy period of one year with a detection probability of 50%. Although calibrated gamma-ray detection systems are capable of determining the concentration of uranium content in NUCPs, it is only in combination with verifiable operator declarations and supporting data, such as flow rate and enrichment, that safeguards conclusions can be drawn.

Semi-annual report on strategic special nuclear material inventory differences

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

Not Available

1978-01-01

This periodic report of Inventory Differences covers the period October 1, 1976, through March 31, 1977 for Department of Energy (DOE) and DOE contractor facilities possessing significant quantities of Strategic Special Nuclear Material (SSNM). Included in this report are the low enriched uranium inventory differences for DOE's gaseous diffusion plant cascades. (LK)

Review of the TREAT Conversion Conceptual Design and Fuel Qualification Plan

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

Diamond, David

The U.S. Department of Energy (DOE) is preparing to re establish the capability to conduct transient testing of nuclear fuels at the Idaho National Laboratory (INL) Transient Reactor Test (TREAT) facility. The original TREAT core went critical in February 1959 and operated for more than 6,000 reactor startups before plant operations were suspended in 1994. DOE is now planning to restart the reactor using the plant's original high-enriched uranium (HEU) fuel. At the same time, the National Nuclear Security Administration (NNSA) Office of Material Management and Minimization Reactor Conversion Program is supporting analyses and fuel fabrication studies that will allowmore » for reactor conversion to low-enriched uranium (LEU) fuel (i.e., fuel with less than 20% by weight 235U content) after plant restart. The TREAT Conversion Program's objectives are to perform the design work necessary to generate an LEU replacement core, to restore the capability to fabricate TREAT fuel element assemblies, and to implement the physical and operational changes required to convert the TREAT facility to use LEU fuel.« less

Benchmark Gamma Spectroscopy Measurements of Uranium Hexafluoride in Aluminmum Pipe with a Sodium Iodide Detector

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

March-Leuba, Jose A; Uckan, Taner; Gunning, John E

2010-01-01

The expected increased demand in fuel for nuclear power plants, combined with the fact that a significant portion of the current supply from the blend down of weapons-source material will soon be coming to an end, has led to the need for new sources of enriched uranium for nuclear fuel. As a result, a number of countries have announced plans, or are currently building, gaseous centrifuge enrichment plants (GCEPs) to supply this material. GCEPs have the potential to produce uranium at enrichments above the level necessary for nuclear fuel purposes-enrichments that make the uranium potentially usable for nuclear weapons. Asmore » a result, there is a critical need to monitor these facilities to ensure that nuclear material is not inappropriately enriched or diverted for unintended use. Significant advances have been made in instrument capability since the current International Atomic Energy Agency (IAEA) monitoring methods were developed. In numerous cases, advances have been made in other fields that have the potential, with modest development, to be applied in safeguards applications at enrichment facilities. A particular example of one of these advances is the flow and enrichment monitor (FEMO). (See Gunning, J. E. et al., 'FEMO: A Flow and Enrichment Monitor for Verifying Compliance with International Safeguards Requirements at a Gas Centrifuge Enrichment Facility,' Proceedings of the 8th International Conference on Facility Operations - Safeguards Interface. Portland, Oregon, March 30-April 4th, 2008.) The FEMO is a conceptual instrument capable of continuously measuring, unattended, the enrichment and mass flow of {sup 235}U in pipes at a GCEP, and consequently increase the probability that the potential production of HEU and/or diversion of fissile material will be detected. The FEMO requires no piping penetrations and can be installed on pipes containing the flow of uranium hexafluoride (UF{sub 6}) at a GCEP. This FEMO consists of separate parts, a flow monitor (FM) and an enrichment monitor (EM). Development of the FM is primarily the responsibility of Oak Ridge National Laboratory, and development of the EM is primarily the responsibility of Los Alamos National Laboratory. The FM will measure {sup 235}U mass flow rate by combining information from measuring the UF{sub 6} volumetric flow rate and the {sup 235}U density. The UF{sub 6} flow rate will be measured using characteristics of the process pumps used in product and tail UF{sub 6} header process lines of many GCEPs, and the {sup 235}U density will be measured using commercially available sodium iodide (NaI) gamma ray scintillation detectors. This report describes the calibration of the portion of the FM that measures the {sup 235}U density. Research has been performed to define a methodology and collect data necessary to perform this calibration without the need for plant declarations. The {sup 235}U density detector is a commercially available system (GammaRad made by Amptek, www.amptek.com) that contains the NaI crystal, photomultiplier tube, signal conditioning electronics, and a multichannel analyzer (MCA). Measurements were made with the detector system installed near four {sup 235}U sources. Two of the sources were made of solid uranium, and the other two were in the form of UF{sub 6} gas in aluminum piping. One of the UF{sub 6} gas sources was located at ORNL and the other at LANL. The ORNL source consisted of two pipe sections (schedule 40 aluminum pipe of 4-inch and 8-inch outside diameter) with 5.36% {sup 235}U enrichment, and the LANL source was a 4-inch schedule 40 aluminum pipe with 3.3% {sup 235}U enrichment. The configurations of the detector on these test sources, as well as on long straight pipe configurations expected to exist at GCEPs, were modeled using the computer code MCNP. The results of the MCNP calculations were used to define geometric correction factors between the test source and the GCEP application. Using these geometric correction factors, the experimental 186 keV counts in the test geometry were extrapolated to the expected GCEP geometry, and calibration curves were developed. A unique method to analyze the measurement was also developed that separated the detector spectrum into the five detectable decay gamma rays emitted by {sup 235}U in the 120 to 200 keV energy range. This analysis facilitated the assignment of a consistent value for the detector counts originating from {sup 235}U decays at 186 keV. This value is also more accurate because it includes the counts from gamma energies other than 186 keV, which results in increased counting statistics for the same measurement time. The 186 keV counts expected as a function of pressure and enrichment are presented in the body of this report. The main result of this research is a calibration factor for 4-inch and 8-inch schedule 40 aluminum pipes. For 4-inch pipes, the {sup 235}U density is 0.62 {sup 235}U g/m{sup 3} per each measured 186 keV count.« less

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

Electrically Heated Testing of the Kilowatt Reactor Using Stirling Technology (KRUSTY) Experiment Using a Depleted Uranium Core

NASA Technical Reports Server (NTRS)

Briggs, Maxwell H.; Gibson, Marc A.; Sanzi, James

2017-01-01

The Kilopower project aims to develop and demonstrate scalable fission-based power technology for systems capable of delivering 110 kW of electric power with a specific power ranging from 2.5 - 6.5 Wkg. This technology could enable high power science missions or could be used to provide surface power for manned missions to the Moon or Mars. NASA has partnered with the Department of Energys National Nuclear Security Administration, Los Alamos National Labs, and Y-12 National Security Complex to develop and test a prototypic reactor and power system using existing facilities and infrastructure. This technology demonstration, referred to as the Kilowatt Reactor Using Stirling TechnologY (KRUSTY), will undergo nuclear ground testing in the summer of 2017 at the Nevada Test Site. The 1 kWe variation of the Kilopower system was chosen for the KRUSTY demonstration. The concept for the 1 kWe flight system consist of a 4 kWt highly enriched Uranium-Molybdenum reactor operating at 800 degrees Celsius coupled to sodium heat pipes. The heat pipes deliver heat to the hot ends of eight 125 W Stirling convertors producing a net electrical output of 1 kW. Waste heat is rejected using titanium-water heat pipes coupled to carbon composite radiator panels. The KRUSTY test, based on this design, uses a prototypic highly enriched uranium-molybdenum core coupled to prototypic sodium heat pipes. The heat pipes transfer heat to two Advanced Stirling Convertors (ASC-E2s) and six thermal simulators, which simulate the thermal draw of full scale power conversion units. Thermal simulators and Stirling engines are gas cooled. The most recent project milestone was the completion of non-nuclear system level testing using an electrically heated depleted uranium (non-fissioning) reactor core simulator. System level testing at the Glenn Research Center (GRC) has validated performance predictions and has demonstrated system level operation and control in a test configuration that replicates the one to be used at the Device Assembly Facility (DAF) at the Nevada National Security Site. Fabrication, assembly, and testing of the depleted uranium core has allowed for higher fidelity system level testing at GRC, and has validated the fabrication methods to be used on the highly enriched uranium core that will supply heat for the DAF KRUSTY demonstration.

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...

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).

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

76 FR 34770 - Rensselaer Polytechnic Institute Critical Experiments Facility; Environmental Assessment and...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-06-14

...)) stainless steel tank with an inner diameter of 2.1 m (7 ft). The reactor is fueled with low enriched uranium SPERT fuel pins. Reactivity control is provided by four Boron-10 control rods. A detailed description of... moderator are stainless steel, thus eliminating the need for routine filtration and demineralization of the...

75 FR 34219 - Revision of Fee Schedules; Fee Recovery for FY 2010

Federal Register 2010, 2011, 2012, 2013, 2014

2010-06-16

....8 $6.3 $7.5 Spent Fuel Storage/Reactor Decommissioning..... -- -- 2.7 0.2 0.2 Test and Research... 2009 fee is also shown for comparative purposes. Table V--Rebaselined Annual Fees FY2009 Annual FY 2010... Decommissioning Test and Research Reactors (Non-power 87,600 81,700 Reactors) High Enriched Uranium Fuel Facility...

10 CFR 110.32 - Information required in an application for a specific license/NRC Form 7.

Code of Federal Regulations, 2011 CFR

2011-01-01

...) and its chemical and physical form. (2) For enriched uranium, the maximum weight percentage of... and its total dollar value. (4) For nuclear reactors, the name of the facility, its design power level..., physical and chemical characteristics, route of transit of shipment, classification (as defined in § 61.55...

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

Federal Register 2010, 2011, 2012, 2013, 2014

2012-02-10

... re-exportation to the end-user, the Japanese customer. The entry met the conditions in the scope of... following the March 11, 2011, earthquake and tsunami that struck Japan, AREVA's Japanese end-use customer was required by the Japanese government to shut down its nuclear power facility pending necessary...

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.

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.

Approach to proliferation risk assessment based on multiple objective analysis framework

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

Andrianov, A.; Kuptsov, I.; Studgorodok 1, Obninsk, Kaluga region, 249030

2013-07-01

The approach to the assessment of proliferation risk using the methods of multi-criteria decision making and multi-objective optimization is presented. The approach allows the taking into account of the specifics features of the national nuclear infrastructure, and possible proliferation strategies (motivations, intentions, and capabilities). 3 examples of applying the approach are shown. First, the approach has been used to evaluate the attractiveness of HEU (high enriched uranium)production scenarios at a clandestine enrichment facility using centrifuge enrichment technology. Secondly, the approach has been applied to assess the attractiveness of scenarios for undeclared production of plutonium or HEU by theft of materialsmore » circulating in nuclear fuel cycle facilities and thermal reactors. Thirdly, the approach has been used to perform a comparative analysis of the structures of developing nuclear power systems based on different types of nuclear fuel cycles, the analysis being based on indicators of proliferation risk.« less

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.

Validation of gamma-ray detection techniques for safeguards monitoring at natural uranium conversion facilities

DOE PAGES

Dewji, Shaheen A.; Lee, Denise L.; Croft, Stephen; ...

2016-03-28

Recent IAEA circulars and policy papers have sought to implement safeguards when any purified aqueous uranium solution or uranium oxides suitable for isotopic enrichment or fuel fabrication exists. Under the revised policy, IAEA Policy Paper 18, the starting point for nuclear material under safeguards was reinterpreted, suggesting that purified uranium compounds should be subject to safeguards procedures no later than the first point in the conversion process. In response to this technical need, a combination of simulation models and experimental measurements were employed to develop and validate concepts of nondestructive assay monitoring systems in a natural uranium conversion plant (NUCP).more » In particular, uranyl nitrate (UO 2(NO 3) 2) solution exiting solvent extraction was identified as a key measurement point (KMP), where gamma-ray spectroscopy was selected as the process monitoring tool. The Uranyl Nitrate Calibration Loop Equipment (UNCLE) facility at Oak Ridge National Laboratory was employed to simulate the full-scale operating conditions of a purified uranium-bearing aqueous stream exiting the solvent extraction process in an NUCP. Nondestructive assay techniques using gamma-ray spectroscopy were evaluated to determine their viability as a technical means for drawing safeguards conclusions at NUCPs, and if the IAEA detection requirements of 1 significant quantity (SQ) can be met in a timely way. This work investigated gamma-ray signatures of uranyl nitrate circulating in the UNCLE facility and evaluated various gamma-ray detector sensitivities to uranyl nitrate. These detector validation activities include assessing detector responses to the uranyl nitrate gamma-ray signatures for spectrometers based on sodium iodide, lanthanum bromide, and high-purity germanium detectors. The results of measurements under static and dynamic operating conditions at concentrations ranging from 10–90 g U/L of natural uranyl nitrate are presented. A range of gamma-ray lines is examined, including attenuation for transmission measurement of density and concentration. It was determined that transmission-corrected gamma-ray spectra provide a reliable way to monitor the 235U concentration of uranyl nitrate solution in transfer pipes in NUCPs. Furthermore, existing predictive and analysis methods are adequate to design and realize practical designs. The 137Cs transmission source employed in this work is viable but not optimal for 235U densitometry determination. Validated simulations assessed the viability of 133Ba and 57Co as alternative densitometry sources. All three gamma-ray detectors are viable for monitoring natural uranium feed; although high-purity germanium is easiest to interpret, it is, however, the least attractive as an installation instrument. Overall, for monitoring throughput in a facility such as UNCLE, emulating the uranium concentration and pump speeds of the Springfields conversion facility in the United Kingdom, an uncertainty of less than 0.17% is required in order to detect the diversion of 1 SQ of uranyl nitrate through changes in uranium concentration over an accountancy period of one year with a detection probability of 50%. As a result, calibrated gamma-ray detection systems are capable of determining the concentration of uranium content in NUCPs, it is only in combination with verifiable operator declarations and supporting data, such as flow rate and enrichment, that safeguards conclusions can be drawn.« less

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

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

Safety Evaluation Report for the Claiborne Enrichment Center, Homer, Louisiana (Docket No. 70-3070)

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

Not Available

1994-01-01

This report documents the US Nuclear Regulatory Commission (NRC) staff review and safety evaluation of the Louisiana Energy Services, L.P. (LES, the applicant) application for a license to possess and use byproduct, source, and special nuclear material and to enrich natural uranium to a maximum of 5 percent U-235 by the gas centrifuge process. The plant, to be known as the Claiborne Enrichment Center (CEC), would be constructed near the town of Homer in Claiborne Parish, Louisiana. At full production in a given year, the plant will receive approximately 4,700 tonnes of feed UF{sub 6} and produce 870 tonnes ofmore » low-enriched UF{sub 6}, and 3,830 tonnes of depleted UF{sub 6} tails. Facility construction, operation, and decommissioning are expected to last 5, 30, and 7 years, respectively. The objective of the review is to evaluate the potential adverse impacts of operation of the facility on worker and public health and safety under both normal operating and accident conditions. The review also considers the management organization, administrative programs, and financial qualifications provided to assure safe design and operation of the facility. The NRC staff concludes that the applicant`s descriptions, specifications, and analyses provide an adequate basis for safety review of facility operations and that construction and operation of the facility does not pose an undue risk to public health and safety.« less

The in-plant evaluation of a uranium NDA system

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

Sprinkle, J.K. Jr.; Baxman, H.R.; Langner, D.G.

1979-12-31

The Los Alamos Scientific Laboratory has an unirradiated enriched uranium reprocessing facility. Various types of solutions are generated in this facility, including distillates and raffinates containing ppm of uranium and concentrated solutions with up to 400 grams U/t. In addition to uranyl nitrate and HNO{sub 3}, the solutions may also contain zirconium, niobium, fluoride, and small amounts of many metals. A uranium solution assay system (USAS) has been installed to allow accurate and more timely process control, accountability, and criticality data to be obtained. The USAS assays are made by a variety of techniques that depend upon state-of-the-art high-resolution Ge(Li)more » gamma-ray spectroscopy integrated with an interactive, user-oriented computer software package. Tight control of the system`s performance is maintained by constantly monitoring the USAS status. Daily measurement control sequences are required, and the user is forced by the software to perform these sequences. Routine assays require 400 or 1000 seconds for a precision of 0.5% over the concentration range of 5--400 g/t. A comparison of the USAS precision and accuracy with that obtained by traditional destructive analytical chemistry techniques (colorimetric and volumetric) is presented.« less

The in-plant evaluation of a uranium NDA system

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

Sprinkle, J.K. Jr.; Baxman, H.R.; Langner, D.G.

1979-01-01

The Los Alamos Scientific Laboratory has an unirradiated enriched uranium reprocessing facility. Various types of solutions are generated in this facility, including distillates and raffinates containing ppm of uranium and concentrated solutions with up to 400 grams U/t. In addition to uranyl nitrate and HNO{sub 3}, the solutions may also contain zirconium, niobium, fluoride, and small amounts of many metals. A uranium solution assay system (USAS) has been installed to allow accurate and more timely process control, accountability, and criticality data to be obtained. The USAS assays are made by a variety of techniques that depend upon state-of-the-art high-resolution Ge(Li)more » gamma-ray spectroscopy integrated with an interactive, user-oriented computer software package. Tight control of the system's performance is maintained by constantly monitoring the USAS status. Daily measurement control sequences are required, and the user is forced by the software to perform these sequences. Routine assays require 400 or 1000 seconds for a precision of 0.5% over the concentration range of 5--400 g/t. A comparison of the USAS precision and accuracy with that obtained by traditional destructive analytical chemistry techniques (colorimetric and volumetric) is presented.« less

Characterization of Uranium Contamination, Transport, and Remediation at Rocky Flats - Across Remediation into Post-Closure

NASA Astrophysics Data System (ADS)

Janecky, D. R.; Boylan, J.; Murrell, M. T.

2009-12-01

The Rocky Flats Site is a former nuclear weapons production facility approximately 16 miles northwest of Denver, Colorado. Built in 1952 and operated by the Atomic Energy Commission and then Department of Energy, the Site was remediated and closed in 2005, and is currently undergoing long-term surveillance and monitoring by the DOE Office of Legacy Management. Areas of contamination resulted from roughly fifty years of operation. Of greatest interest, surface soils were contaminated with plutonium, americium, and uranium; groundwater was contaminated with chlorinated solvents, uranium, and nitrates; and surface waters, as recipients of runoff and shallow groundwater discharge, have been contaminated by transport from both regimes. A region of economic mineralization that has been referred to as the Colorado Mineral Belt is nearby, and the Schwartzwalder uranium mine is approximately five miles upgradient of the Site. Background uranium concentrations are therefore elevated in many areas. Weapons-related activities included work with enriched and depleted uranium, contributing anthropogenic content to the environment. Using high-resolution isotopic analyses, Site-related contamination can be distinguished from natural uranium in water samples. This has been instrumental in defining remedy components, and long-term monitoring and surveillance strategies. Rocky Flats hydrology interlinks surface waters and shallow groundwater (which is very limited in volume and vertical and horizontal extent). Surface water transport pathways include several streams, constructed ponds, and facility surfaces. Shallow groundwater has no demonstrated connection to deep aquifers, and includes natural preferential pathways resulting primarily from porosity in the Rocky Flats alluvium, weathered bedrock, and discontinuous sandstones. In addition, building footings, drains, trenches, and remedial systems provide pathways for transport at the site. Removal of impermeable surfaces (buildings, roads, and so on) during the Site closure efforts resulted in major changes to surface and shallow groundwater flow. Consistent with previous documentation of uranium operations and contamination, only very small amounts of highly enriched uranium are found in a small number of water samples, generally from the former Solar Ponds complex and central Industrial Area. Depleted uranium is more widely distributed at the site, and water samples exhibit the full range of depleted plus natural uranium mixtures. However, one third of the samples are found to contain only natural uranium, and three quarters of the samples are found to contain more than 90% natural uranium - substantial fractions given that the focus of these analyses was on evaluating potentially contaminated waters. Following site closure, uranium concentrations have increased at some locations, particularly for surface water samples. Overall, isotopic ratios at individual locations have been relatively consistent, indicating that the increases in concentrations are due to decreases in dilution flow following removal of impermeable surfaces and buildings.

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/

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...

Calculated criticality for sup 235 U/graphite systems using the VIM Monte Carlo code

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

Collins, P.J.; Grasseschi, G.L.; Olsen, D.N.

1992-01-01

Calculations for highly enriched uranium and graphite systems gained renewed interest recently for the new production modular high-temperature gas-cooled reactor (MHTGR). Experiments to validate the physics calculations for these systems are being prepared for the Transient Reactor Test Facility (TREAT) reactor at Argonne National Laboratory (ANL-West) and in the Compact Nuclear Power Source facility at Los Alamos National Laboratory. The continuous-energy Monte Carlo code VIM, or equivalently the MCNP code, can utilize fully detailed models of the MHTGR and serve as benchmarks for the approximate multigroup methods necessary in full reactor calculations. Validation of these codes and their associated nuclearmore » data did not exist for highly enriched {sup 235}U/graphite systems. Experimental data, used in development of more approximate methods, dates back to the 1960s. The authors have selected two independent sets of experiments for calculation with the VIM code. The carbon-to-uranium (C/U) ratios encompass the range of 2,000, representative of the new production MHTGR, to the ratio of 10,000 in the fuel of TREAT. Calculations used the ENDF/B-V data.« less

Is It Time To Consider Global Sharing of Integral Physics Data?

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

Harold F. McFarlane

The innocent days of the Atoms for Peace program vanished with the suicide attack on the World Trade Center in New York City that occurred while the GLOBAL 2001 international nuclear fuel cycle conference was convened in Paris. Today’s reality is that maintaining an inventory of unirradiated highly enriched uranium or plutonium for critical experiments requires a facility to accept substantial security cost and intrusion. In the context of a large collection of benchmark integral experiments collected over several decades and the ongoing rapid advances in computer modeling and simulation, there seems to be ample incentive to reduce both themore » number of facilities and material inventory quantities worldwide. As a result of ongoing nonproliferation initiatives, there are viable programs that will accept highly enriched uranium for down blending into commercial fuel. Nevertheless, there are formidable hurdles to overcome before national institutions will voluntarily give up existing nuclear research capabilities. GLOBAL 2005 was the appropriate forum to begin fostering a new spirit of cooperation that could lead to improved international security and better use of precious research and development resources, while ensuring access to existing and future critical experiment data.« less

The Mailbox Computer System for the IAEA verification experiment on HEU downlending at the Portsmouth Gaseous Diffusion Plant

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

Aronson, A.L.; Gordon, D.M.

IN APRIL 1996, THE UNITED STATES (US) ADDED THE PORTSMOUTH GASEOUS DIFFUSION PLANT TO THE LIST OF FACILITIES ELIGIBLE FOR THE APPLICATION OF INTERNATIONAL ATOMIC ENERGY AGENCY (IAEA) SAFEGUARDS. AT THAT TIME, THE US PROPOSED THAT THE IAEA CARRY OUT A ''VERIFICATION EXPERIMENT'' AT THE PLANT WITH RESPECT TO DOOWNBLENDING OF ABOUT 13 METRIC TONS OF HIGHLY ENRICHED URANIUM (HEU) IN THE FORM OF URANIUM HEXAFLUROIDE (UF6). DURING THE PERIOD DECEMBER 1997 THROUGH JULY 1998, THE IAEA CARRIED OUT THE REQUESTED VERIFICATION EXPERIMENT. THE VERIFICATION APPROACH USED FOR THIS EXPERIMENT INCLUDED, AMONG OTHER MEASURES, THE ENTRY OF PROCESS-OPERATIONAL DATA BYmore » THE FACILITY OPERATOR ON A NEAR-REAL-TIME BASIS INTO A ''MAILBOX'' COMPUTER LOCATED WITHIN A TAMPER-INDICATING ENCLOSURE SEALED BY THE IAEA.« 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

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

Development of solid materials for UF 6 sampling: FY16 Annual Report

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

Smith, Nicholas; Savina, Joseph; Hebden, Andrew

2016-10-31

A handheld implementation of the ABACC-developed Cristallini method, which captures uranium hexafluoride samples as an inert salt, was organized in FY17 and succeeded in demonstrating the handheld sampler concept with reactive hexafluoride gases. The Cristallini method relies on the use of a hydrated substrate to react the incoming hexafluoride resulting in the formation of a stable uranyl fluoride salt. The Cristallini method has been demonstrated as a facility modification installed near the sampling tap of a gas centrifuge enrichment plant. While very successful in reducing the hazards of uranium hexafluoride sample, the method still takes a considerable amount of timemore » and can only be used in facilities where the apparatus has been installed; this arrangement generally prohibits the sampling of filled cylinders that have already exited the facility and have been deposited in the on-site tank storage yard. The handheld unit under development will allow the use of the Cristallini method at facilities that have not been converted as well as tanks in the storage yard. The handheld system utilizes an active vacuum system, rather than a passive vacuum system in the facility setup, to drive the uranium hexafluoride onto the adsorbing media. The handheld unit will be battery operated for fully autonomous operation and will include onboard pressure sensing and flushing capability. To date, the system concept of operations was demonstrated with tungsten hexafluoride that showed the active vacuum pump with multiple cartridges of adsorbing media was viable. Concurrently, the hardened prototype system was developed and tested; removable sample cartridges were developed (the only non-COTS component to date); and preparations were made for uranium tests and a domestic field test.« less

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)

Basis for Interim Operation for Fuel Supply Shutdown Facility

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

BENECKE, M.W.

2003-02-03

This document establishes the Basis for Interim Operation (BIO) for the Fuel Supply Shutdown Facility (FSS) as managed by the 300 Area Deactivation Project (300 ADP) organization in accordance with the requirements of the Project Hanford Management Contract procedure (PHMC) HNF-PRO-700, ''Safety Analysis and Technical Safety Requirements''. A hazard classification (Benecke 2003a) has been prepared for the facility in accordance with DOE-STD-1027-92 resulting in the assignment of Hazard Category 3 for FSS Facility buildings that store N Reactor fuel materials (303-B, 3712, and 3716). All others are designated Industrial buildings. It is concluded that the risks associated with the currentmore » and planned operational mode of the FSS Facility (uranium storage, uranium repackaging and shipment, cleanup, and transition activities, etc.) are acceptable. The potential radiological dose and toxicological consequences for a range of credible uranium storage building have been analyzed using Hanford accepted methods. Risk Class designations are summarized for representative events in Table 1.6-1. Mitigation was not considered for any event except the random fire event that exceeds predicted consequences based on existing source and combustible loading because of an inadvertent increase in combustible loading. For that event, a housekeeping program to manage transient combustibles is credited to reduce the probability. An additional administrative control is established to protect assumptions regarding source term by limiting inventories of fuel and combustible materials. Another is established to maintain the criticality safety program. Additional defense-in-depth controls are established to perform fire protection system testing, inspection, and maintenance to ensure predicted availability of those systems, and to maintain the radiological control program. It is also concluded that because an accidental nuclear criticality is not credible based on the low uranium enrichment, the form of the uranium, and the required controls, a Criticality Alarm System (CAS) is not required as allowed by DOE Order 420.1 (DOE 2000).« less

11. AERIAL VIEW LOOKING NORTH AT THE BUILDING 800 ...

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

11. AERIAL VIEW LOOKING NORTH AT THE BUILDING 800 - AREA COMPLEX. ENRICHED URANIUM COMPONENTS WERE MANUFACTURED IN THIS AREA OF THE SITE. BUILDING 881, IN THE RIGHT FOREGROUND OF THE PHOTOGRAPH, WAS THE ORIGINAL PLANT B. BUILDING 883, USED FOR ROLLING AND FORMING URANIUM COMPONENTS, IS DIRECTLY TO THE NORTH OF BUILDING 881. TO THE EAST OF BUILDING 883 IS BUILDING 885, A RESEARCH AND DEVELOPMENT FACILITY FOR ALLOYS AND NON-PLUTONIUM METALS. IN THE FOREGROUND TO THE WEST OF BUILDING 881 IS AN OFFICE BUILDING, 850 (6/7/90). - Rocky Flats Plant, Bounded by Indiana Street & Routes 93, 128 & 72, Golden, Jefferson County, CO

Utilization of the Philippine Research Reactor as a training facility for nuclear power plant operators

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

Palabrica, R.J.

1981-01-01

The Philippines has a 1-MW swimming-pool reactor facility operated by the Philippine Atomic Energy Commission (PAEC). The reactor is light-water moderated and cooled, graphite reflected, and fueled with 90% enriched uranium. Since it became critical in 1963 it has been utilized for research, radioisotope production, and training. It was used initially in the training of PAEC personnel and other research institutions and universities. During the last few years, however, it has played a key role in training personnel for the Philippine Nuclear Power Project (PNPP).

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...

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

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

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

The United States, Russian Federation, and the International Atomic Energy Agency have been working together on a program called the Russian Research Reactor Fuel Return (RRRFR) Program, which is part of the Global Threat Reduction Initiative. The purpose of this program is to return Soviet or Russian-supplied high-enriched uranium (HEU) fuel, currently stored at Russian-designed research reactors throughout the world, to Russia. In February 2003, the RRRFR Program began discussions with the Nuclear Research Institute (NRI) in Rež, Czech Republic, about returning their HEU spent nuclear fuel to the Russian Federation for reprocessing. In March 2005, the U.S. Department ofmore » Energy signed a contract with NRI to perform all activities needed for transporting their HEU spent nuclear fuel to Russia. After 2 years of intense planning, preparations, and coordination at NRI and with three other countries, numerous organizations and agencies, and a Russian facility, this shipment is scheduled for completion before the end of 2007. This paper will provide a summary of activities completed for making this international shipment. This paper contains an introduction and background of the RRRFR Program and the NRI shipment project. It summarizes activities completed in preparation for the shipment, including facility preparations at NRI in Rež and FSUE “Mayak” in Ozyorsk, Russia; a new transportation cask system; regulatory approvals; transportation planning and preparation in the Czech Republic, Slovakia, Ukraine, and the Russian Federation though completion of the Unified Project and Special Ecological Programs. The paper also describes fuel loading and cask preparations at NRI and final preparations/approvals for transporting the shipment across the Czech Republic, Slovakia, Ukraine, and the Russian Federation to FSUE Mayak where the HEU spent nuclear fuel will be processed, the uranium will be downblended and made into low-enriched uranium fuel for commercial reactor use, and the high-level waste from the processing will be stabilized and stored for less than 20 years before being sent back to the Czech Republic for final disposition. Finally, the paper contains a section for the summary and conclusions.« less

Nuclear Energy Policy

DTIC Science & Technology

2008-01-28

2007. Requires commercial nuclear power plants to transfer spent fuel from pools to dry storage casks and then convey title to the Secretary of Energy...far more economical options for reducing fossil fuel use .15 (For more on federal incentives and the economics of nuclear power, see CRS Report RL33442...uranium enrichment, spent fuel recycling (also called reprocessing), and other fuel cycle facilities that could be used to produce nuclear weapons

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

Enrichment Assay Methods Development for the Integrated Cylinder Verification System

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

Smith, Leon E.; Misner, Alex C.; Hatchell, Brian K.

2009-10-22

International Atomic Energy Agency (IAEA) inspectors currently perform periodic inspections at uranium enrichment plants to verify UF6 cylinder enrichment declarations. Measurements are typically performed with handheld high-resolution sensors on a sampling of cylinders taken to be representative of the facility's entire product-cylinder inventory. Pacific Northwest National Laboratory (PNNL) is developing a concept to automate the verification of enrichment plant cylinders to enable 100 percent product-cylinder verification and potentially, mass-balance calculations on the facility as a whole (by also measuring feed and tails cylinders). The Integrated Cylinder Verification System (ICVS) could be located at key measurement points to positively identify eachmore » cylinder, measure its mass and enrichment, store the collected data in a secure database, and maintain continuity of knowledge on measured cylinders until IAEA inspector arrival. The three main objectives of this FY09 project are summarized here and described in more detail in the report: (1) Develop a preliminary design for a prototype NDA system, (2) Refine PNNL's MCNP models of the NDA system, and (3) Procure and test key pulse-processing components. Progress against these tasks to date, and next steps, are discussed.« less

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

Viability Study for an Unattended UF 6 Cylinder Verification Station: Phase I Final Report

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

Smith, Leon E.; Miller, Karen A.; Garner, James R.

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) that could provide automated, independent verification of the declared relative enrichment, 235U mass, total uranium mass and identification for all declared UF 6 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 US Support Program team consisted of Pacific Northwest National Laboratory (PNNL, lead), Los Alamos National Laboratory (LANL), Oak Ridge National Laboratory (ORNL) and Savanah River National Laboratory (SRNL). At the core of the viability study is a long-term field trial of a prototype UCVS system at a Westinghouse fuel fabrication facility. A key outcome of the study is 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 report provides context for the UCVS concept and the field trial: potential UCVS implementation concepts at an enrichment facility; an overview of UCVS prototype design; field trial objectives and activities. Field trial 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 the contents of a given cylinder are consistent with previous scans. A modeling study, combined with field-measured instrument uncertainties, provides an assessment of the partial-defect sensitivity of HEVA and PNEM for both one-time assay and (repeated) NDA Fingerprint verification scenarios. The findings presented in this report represent a significant step forward in the community’s understanding of the strengths and limitations of the PNEM and HEVA NDA methods, and the viability of the UCVS concept in front-end fuel cycle facilities. This experience will inform Phase II of the UCVS viability study, should the IAEA pursue it.« less

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

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

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

Benton, J; Wall, D; Parker, E

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

Performance and Fabrication Status of TREAT LEU Conversion Conceptual Design Concepts

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

IJ van Rooyen; SR Morrell; AE Wright

2014-10-01

Resumption of transient testing at the TREAT facility was approved in February 2014 to meet U.S. Department of Energy (DOE) objectives. The National Nuclear Security Administration’s Global Threat Reduction Initiative Convert Program is evaluating conversion of TREAT from its existing highly enriched uranium (HEU) core to a new core containing low enriched uranium (LEU). This paper describes briefly the initial pre-conceptual designs screening decisions with more detailed discussions on current feasibility, qualification and fabrication approaches. Feasible fabrication will be shown for a LEU fuel element assembly that can meet TREAT design, performance, and safety requirements. The statement of feasibility recognizesmore » that further development, analysis, and testing must be completed to refine the conceptual design. Engineering challenges such as cladding oxidation, high temperature material properties, and fuel block fabrication along with neutronics performance, will be highlighted. Preliminary engineering and supply chain evaluation provided confidence that the conceptual designs can be achieved.« less

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.

Nominations for the 2017 NNSA Pollution Prevention Awards

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

Salzman, Sonja L.; Ballesteros Rodriguez, Sonia; Lopez, Lorraine Bonds

In the field of nuclear forensics, one of the biggest challenges is to dissolve postdetonation debris for analysis. Debris generated after a nuclear detonation is a glassy material that is difficult to dissolve with chemicals. Traditionally, concentrated nitric acid, hydrofluoric acid, or sulfuric acid are employed during the dissolution. These acids, due to their corrosive nature, are not suitable for in-field/on-site sample preparations. Uranium oxides are commonly present in nuclear fuel processing plants and nuclear research facilities. In uranium oxides, the level of uranium isotope enrichment is a sensitive indicator for nuclear nonproliferation and is monitored closely by the Internationalmore » Atomic Energy Agency (IAEA) to ensure there is no misuse of nuclear material or technology for nuclear weapons. During an IAEA on-site inspection at a facility, environmental surface swipe samples are collected and transported to the IAEA headquarters or network of analytical laboratories for further processing. Uranium oxide particles collected on the swipe medium are typically dissolved with inorganic acids and are then analyzed for uranium isotopic compositions. To improve the responsiveness of on-site inspections, in-field detection techniques have been recently explored. However, in-field analysis is bottlenecked by time-consuming and hazardous dissolution procedures, as corrosive inorganic acids must be used. Corrosive chemicals are difficult to use in the field due to personnel safety considerations, and the transportation of such chemicals is highly regulated. It was therefore necessary to develop fast uranium oxide dissolution methods using less hazardous chemicals in support of the rapid infield detection of anomalies in declared nuclear processes.« less

Nuclear criticality safety assessment of the low level radioactive waste disposal facility trenches

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

Kahook, S.D.

1994-04-01

Results of the analyses performed to evaluate the possibility of nuclear criticality in the Low Level Radioactive Waste Disposal Facility (LLRWDF) trenches are documented in this report. The studies presented in this document are limited to assessment of the possibility of criticality due to existing conditions in the LLRWDF. This document does not propose nor set limits for enriched uranium (EU) burial in the LLRWDF and is not a nuclear criticality safety evaluation nor analysis. The calculations presented in the report are Level 2 calculations as defined by the E7 Procedure 2.31, Engineering Calculations.

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...

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.

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.

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

Fast Neutron Spectrum Potassium Worth for Space Power Reactor Design Validation

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

Bess, John D.; Marshall, Margaret A.; Briggs, J. Blair

2015-03-01

A variety of critical experiments were constructed of enriched uranium metal (oralloy ) during the 1960s and 1970s at the Oak Ridge Critical Experiments Facility (ORCEF) in support of criticality safety operations at the Y-12 Plant. The purposes of these experiments included the evaluation of storage, casting, and handling limits for the Y-12 Plant and providing data for verification of calculation methods and cross-sections for nuclear criticality safety applications. These included solid cylinders of various diameters, annuli of various inner and outer diameters, two and three interacting cylinders of various diameters, and graphite and polyethylene reflected cylinders and annuli. Ofmore » the hundreds of delayed critical experiments, one was performed that consisted of uranium metal annuli surrounding a potassium-filled, stainless steel can. The outer diameter of the annuli was approximately 13 inches (33.02 cm) with an inner diameter of 7 inches (17.78 cm). The diameter of the stainless steel can was 7 inches (17.78 cm). The critical height of the configurations was approximately 5.6 inches (14.224 cm). The uranium annulus consisted of multiple stacked rings, each with radial thicknesses of 1 inch (2.54 cm) and varying heights. A companion measurement was performed using empty stainless steel cans; the primary purpose of these experiments was to test the fast neutron cross sections of potassium as it was a candidate for coolant in some early space power reactor designs.The experimental measurements were performed on July 11, 1963, by J. T. Mihalczo and M. S. Wyatt (Ref. 1) with additional information in its corresponding logbook. Unreflected and unmoderated experiments with the same set of highly enriched uranium metal parts were performed at the Oak Ridge Critical Experiments Facility in the 1960s and are evaluated in the International Handbook for Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook) with the identifier HEU MET FAST 051. Thin graphite reflected (2 inches or less) experiments also using the same set of highly enriched uranium metal parts are evaluated in HEU MET FAST 071. Polyethylene-reflected configurations are evaluated in HEU-MET-FAST-076. A stack of highly enriched metal discs with a thick beryllium top reflector is evaluated in HEU-MET-FAST-069, and two additional highly enriched uranium annuli with beryllium cores are evaluated in HEU-MET-FAST-059. Both detailed and simplified model specifications are provided in this evaluation. Both of these fast neutron spectra assemblies were determined to be acceptable benchmark experiments. The calculated eigenvalues for both the detailed and the simple benchmark models are within ~0.26 % of the benchmark values for Configuration 1 (calculations performed using MCNP6 with ENDF/B-VII.1 neutron cross section data), but under-calculate the benchmark values by ~7s because the uncertainty in the benchmark is very small: ~0.0004 (1s); for Configuration 2, the under-calculation is ~0.31 % and ~8s. Comparison of detailed and simple model calculations for the potassium worth measurement and potassium mass coefficient yield results approximately 70 – 80 % lower (~6s to 10s) than the benchmark values for the various nuclear data libraries utilized. Both the potassium worth and mass coefficient are also deemed to be acceptable benchmark experiment measurements.« less

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.

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

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

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

2003-02-27

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

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

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

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

Estimate of radiation release from MIT reactor with un-finned LEU core during Maximum Hypothetical Accident

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

Sun, Kaichao; Hu, Lin-wen; Newton, Thomas

2017-05-01

The Massachusetts Institute of Technology Reactor (MITR-II) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. At 6 MW, it delivers neutron flux and energy spectrum comparable to light water reactor (LWR) power reactors in a compact core using highly enriched uranium (HEU) fuel. In the framework of nonproliferation policy, the international community aims to minimize the use of HEU in civilian facilities. Within this context, research and test reactors have started a program to convert HEU fuel to low enriched uranium (LEU) fuel. A new type of LEU fuel basedmore » on a high density alloy of uranium and molybdenum (U-10Mo) is expected to allow the conversion of U.S. domestic high performance reactors like MITR. The current study focuses on the impacts of MITR Maximum Hypothetical Accident (MHA), which is also the Design Basis Accident (DBA), with LEU fuel. The MHA for the MITR is postulated to be a coolant flow blockage in the fuel element that contains the hottest fuel plate. It is assumed that the entire active portion of five fuel plates melts. The analysis shows that, within a 2-h period and by considering all the possible radiation sources and dose pathways, the overall off-site dose is 302.1 mrem (1 rem ¼ 0.01 Sv) Total Effective Dose Equivalent (TEDE) at 8 m exclusion area boundary (EAB) and a higher dose of 392.8 mrem TEDE is found at 21 m EAB. In all cases the dose remains below the 500 mrem total TEDE limit goal based on NUREG-1537 guidelines.« less

RESULTS FROM A DEMONSTRATION OF RF-BASED UF6 CYLINDER ACCOUNTING AND TRACKING SYSTEM INSTALLED AT A USEC FACILITY

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

Pickett, Chris A; Kovacic, Donald N; Morgan, Jim

Approved industry-standard cylinders are used globally for storing and transporting uranium hexafluoride (UF{sub 6}) at uranium enrichment plants and processing facilities. To verify that no diversion or undeclared production of nuclear material involving UF{sub 6} cylinders at the facility has occurred, the International Atomic Energy Agency (IAEA) conducts periodic, labor-intensive physical inspections to validate facility records, cylinder identities, and cylinder weights. A reliable cylinder monitoring system that would improve overall inspector effectiveness would be a significant improvement to the current international safeguards inspection regime. Such a system could include real-time unattended monitoring of cylinder movements, situation-specific rules-based event detection algorithms,more » and the capability to integrate with other types of safeguards technologies. This type of system could provide timely detection of abnormal operational activities that may be used to ensure more appropriate and efficient responses by the IAEA. A system of this type can reduce the reliance on paper records and have the additional benefit of facilitating domestic safeguards at the facilities at which it is installed. A radio-frequency (RF)-based system designed to track uranium hexafluoride (UF{sub 6}) cylinders during processing operations was designed, assembled, and tested at the United States Enrichment Corporation (USEC) facility in Portsmouth, Ohio, to determine the operational feasibility and durability of RF technology. The overall objective of the effort was to validate the robustness of RF technology for potential use as a future international safeguards tool for tracking UF6 cylinders at uranium-processing facilities. The results to date indicate that RF tags represent a feasible technique for tracking UF{sub 6} cylinders in operating facilities. Additional work will be needed to improve the operational robustness of the tags for repeated autoclave processing and to add tamper-indicating and data authentication features to some of the pertinent system components. Future efforts will focus on these needs along with implementing protocols relevant to IAEA safeguards. The work detailed in this report demonstrates the feasibility of constructing RF devices that can survive the operational rigors associated with the transportation, storage, and processing of UF6 cylinders. The system software specially designed for this project is called Cylinder Accounting and Tracking System (CATS). This report details the elements of the CATS rules-based architecture and its use in safeguards-monitoring and asset-tracking applications. Information is also provided on improvements needed to make the technology ready, as well as options for improving the safeguards aspects of the technology. The report also includes feedback from personnel involved in the testing, as well as individuals who could utilize an RF-based system in supporting the performance of their work. The system software was set up to support a Mailbox declaration, where a declaration can be made either before or after cylinder movements take place. When the declaration is made before cylinders move, the operators must enter this information into CATS. If the IAEA then shows up unexpectedly at the facility, they can see how closely the operational condition matches the declaration. If the declaration is made after the cylinders move, this provides greater operational flexibility when schedules are interrupted or are changed, by allowing operators to declare what moves have been completed. The IAEA can then compare where cylinders are with where CATS or the system says they are located. The ability of CATS to automatically generate Mailbox declarations is seen by the authors as a desirable feature. The Mailbox approach is accepted by the IAEA but has not been widely implemented (and never in enrichment facilities). During the course of this project, we have incorporated alternative methods for implementation.« less

Refurbishment of uranium hexafluoride cylinder storage yards C-745-K, L, M, N, and P and construction of a new uranium hexafluoride cylinder storage yard (C-745-T) at the Paducah Gaseous Diffusion Plant, Paducah, Kentucky

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

NONE

1996-07-01

The Paducah Gaseous Diffusion Plant (PGDP) is a uranium enrichment facility owned by the US Department of Energy (DOE). A residual of the uranium enrichment process is depleted uranium hexafluoride (UF6). Depleted UF6, a solid at ambient temperature, is stored in 32,200 steel cylinders that hold a maximum of 14 tons each. Storage conditions are suboptimal and have resulted in accelerated corrosion of cylinders, increasing the potential for a release of hazardous substances. Consequently, the DOE is proposing refurbishment of certain existing yards and construction of a new storage yard. This environmental assessment (EA) evaluates the impacts of the proposedmore » action and no action and considers alternate sites for the proposed new storage yard. The proposed action includes (1) renovating five existing cylinder yards; (2) constructing a new UF6 storage yard; handling and onsite transport of cylinders among existing yards to accommodate construction; and (4) after refurbishment and construction, restacking of cylinders to meet spacing and inspection requirements. Based on the results of the analysis reported in the EA, DOE has determined that the proposed action is not a major Federal action that would significantly affect the quality of the human environment within the context of the National Environmental Policy Act of 1969. Therefore, DOE is issuing a Finding of No Significant Impact. Additionally, it is reported in this EA that the loss of less than one acre of wetlands at the proposed project site would not be a significant adverse impact.« less

Central Asia: Regional Developments and Implications for U.S. Interests

DTIC Science & Technology

2009-05-28

Organization (SCO; see above, “Obstacles to Peace and Independence: Regional Tensions and Conflicts”) that stated that “as large - scale military...producers of low enriched uranium. Kazakhstan had a fast breeder reactor at Aktau that was the world’s only nuclear desalinization facility. In 1997 and...88 All the states of the region possess large - scale resources that could contribute to the region becoming a “new silk road” of trade and commerce

The Forgotten Service: Determining the US Army’s Role in Shaping American Strategy in the Asia-Pacific Region

DTIC Science & Technology

2014-06-01

The Limits of Air Power: The American Bombing of North Vietnam (Lincoln: University of Nebraska Press, 2006, 43. 93 Russell F. Weigley, The...Korea’s bombardment of Yeonpyong, its construction of new uranium enrichment facilities, and development of nuclear warheads and missiles.23 The...89. Weigley, Russell F. The American Way of War: A History of United States Military Strategy and Policy. Bloomington: Indiana University Press

MCNP6 simulated performance of Micro-Pocket Fission Detectors (MPFDs) in the Transient REActor Test (TREAT) Facility

DOE PAGES

Reichenberger, Michael A.; Patel, Vishal K.; Roberts, Jeremy A.; ...

2017-03-03

Here, Micro-Pocket Fission Detectors (MPFDs) are under development for in-core neutron flux measurements at the Transient REActor Test facility (TREAT) and in other experiments at Idaho National Laboratory (INL). The sensitivity of MPFDs to the energy dependent neutron flux at TREAT has been determined for 0.0300-μm thick active material coatings of 242Pu, 232Th, natural uranium, and 93% enriched 235U. Self-shielding effects in the active material of the MPFD was also confirmed to be negligible. Finally, fission fragment energy deposition was found to be in conformance with previously reported results.

Performance testing accountability measurements

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

Oldham, R.D.; Mitchell, W.G.; Spaletto, M.I.

The New Brunswick Laboratory (NBL) provides assessment support to the DOE Operations Offices in the area of Material Control and Accountability (MC and A). During surveys of facilities, the Operations Offices have begun to request from NBL either assistance in providing materials for performance testing of accountability measurements or both materials and personnel to do performance testing. To meet these needs, NBL has developed measurement and measurement control performance test procedures and materials. The present NBL repertoire of performance tests include the following: (1) mass measurement performance testing procedures using calibrated and traceable test weights, (2) uranium elemental concentration (assay)more » measurement performance tests which use ampulated solutions of normal uranyl nitrate containing approximately 7 milligrams of uranium per gram of solution, and (3) uranium isotopic measurement performance tests which use ampulated uranyl nitrate solutions with enrichments ranging from 4% to 90% U-235. The preparation, characterization, and packaging of the uranium isotopic and assay performance test materials were done in cooperation with the NBL Safeguards Measurements Evaluation Program since these materials can be used for both purposes.« less

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)

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.

Neutronics Analyses of the Minimum Original HEU TREAT Core

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

Kontogeorgakos, D.; Connaway, H.; Yesilyurt, G.

2014-04-01

This work was performed to support the feasibility study on the potential conversion of the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory from the use of high-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by the GTRI Reactor Conversion staff at the Argonne National Laboratory (ANL). The objective of this study was to validate the MCNP model of the TREAT reactor with the well-documented measurements which were taken during the start-up and early operation of TREAT. Furthermore, the effect of carbon graphitization was also addressed. The graphitization level was assumedmore » to be 100% (ANL/GTRI/TM-13/4). For this purpose, a set of experiments was chosen to validate the TREAT MCNP model, involving the approach to criticality procedure, in-core neutron flux measurements with foils, and isothermal temperature coefficient and temperature distribution measurements. The results of this study extended the knowledge base for the TREAT MCNP calculations and established the credibility of the MCNP model to be used in the core conversion feasibility analysis.« less

Pre-conceptual Development and characterization of an extruded graphite composite fuel for the TREAT Reactor

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

Luther, Erik; Rooyen, Isabella van; Leckie, Rafael

2015-03-01

In an effort to explore fuel systems that are more robust under accident scenarios, the DOE-NE has identified the need to resume transient testing. The Transient Reactor Test (TREAT) facility has been identified as the preferred option for the resumption of transient testing of nuclear fuel in the United States. In parallel, NNSA’s Global Threat Reduction Initiative (GTRI) Convert program is exploring the needs to replace the existing highly enriched uranium (HEU) core with low enriched uranium (LEU) core. In order to construct a new LEU core, materials and fabrication processes similar to those used in the initial core fabricationmore » must be identified, developed and characterized. In this research, graphite matrix fuel blocks were extruded and materials properties of were measured. Initially the extrusion process followed the historic route; however, the project was expanded to explore methods to increase the graphite content of the fuel blocks and explore modern resins. Materials properties relevant to fuel performance including density, heat capacity and thermal diffusivity were measured. The relationship between process defects and materials properties will be discussed.« less

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

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

Pope, M. A.; DeHart, M. D.; Morrell, S. R.

2015-03-01

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

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

Marshall, Margaret A.

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 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...

The D-D Neutron Generator as an Alternative to Am(Li) Isotopic Neutron Source in the Active Well Coincidence Counter

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

McElroy, Robert Dennis; Cleveland, Steven L.

The 235U mass assay of bulk uranium items, such as oxide canisters, fuel pellets, and fuel assemblies, is not achievable by traditional gamma-ray assay techniques due to the limited penetration of the item by the characteristic 235U gamma rays. Instead, fast neutron interrogation methods such as active neutron coincidence counting must be used. For international safeguards applications, the most commonly used active neutron systems, the Active Well Coincidence Counter (AWCC), Uranium Neutron Collar (UNCL) and 252Cf Shuffler, rely on fast neutron interrogation using an isotopic neutron source [i.e., 252Cf or Am(Li)] to achieve better measurement accuracies than are possible usingmore » gamma-ray techniques for high-mass, high-density items. However, the Am(Li) sources required for the AWCC and UNCL systems are no longer manufactured, and newly produced systems rely on limited supplies of sources salvaged from disused instruments. The 252Cf shuffler systems rely on the use of high-output 252Cf sources, which while still available have become extremely costly for use in routine operations and require replacement every five to seven years. Lack of a suitable alternative neutron interrogation source would leave a potentially significant gap in the safeguarding of uranium processing facilities. In this work, we made use of Oak Ridge National Laboratory’s (ORNL’s) Large Volume Active Well Coincidence Counter (LV-AWCC) and a commercially available deuterium-deuterium (D-D) neutron generator to examine the potential of the D-D neutron generator as an alternative to the isotopic sources. We present the performance of the LV-AWCC with D-D generator for the assay of 235U based on the results of Monte Carlo N-Particle (MCNP) simulations and measurements of depleted uranium (DU), low enriched uranium (LEU), and highly enriched uranium (HEU) items.« less

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,...

Decommissioning ALARA programs Cintichem decommissioning experience

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

Adler, J.J.; LaGuardia, T.S.

1995-03-01

The Cintichem facility, originally the Union Carbide Nuclear Company (UCNC) Research Center, consisted primarily of a 5MW pool type reactor linked via a four-foot-wide by twelve-foot-deep water-filled canal to a bank of five adjacent hot cells. Shortly after going into operations in the early 1960s, the facility`s operations expanded to provide various reactor-based products and services to a multitude of research, production, medical, and education groups. From 1968 through 1972, the facility developed a process of separating isotopes from mixed fission products generated by irradiating enriched Uranium target capsules. By the late 1970s, 20 to 30 capsules were being processedmore » weekly, with about 200,000 curies being produced per week. Several isotopes such as Mo{sup 99}, I{sup 131}, and Xe{sup 133} were being extracted for medical use.« less

Uranium-Loaded Water Treatment Resins: 'Equivalent Feed' at NRC and Agreement State-Licensed Uranium Recovery Facilities - 12094

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

Camper, Larry W.; Michalak, Paul; Cohen, Stephen

Community Water Systems (CWSs) are required to remove uranium from drinking water to meet EPA standards. Similarly, mining operations are required to remove uranium from their dewatering discharges to meet permitted surface water discharge limits. Ion exchange (IX) is the primary treatment strategy used by these operations, which loads uranium onto resin beads. Presently, uranium-loaded resin from CWSs and mining operations can be disposed as a waste product or processed by NRC- or Agreement State-licensed uranium recovery facilities if that licensed facility has applied for and received permission to process 'alternate feed'. The disposal of uranium-loaded resin is costly andmore » the cost to amend a uranium recovery license to accept alternate feed can be a strong disincentive to commercial uranium recovery facilities. In response to this issue, the NRC issued a Regulatory Issue Summary (RIS) to clarify the agency's policy that uranium-loaded resin from CWSs and mining operations can be processed by NRC- or Agreement State-licensed uranium recovery facilities without the need for an alternate feed license amendment when these resins are essentially the same, chemically and physically, to resins that licensed uranium recovery facilities currently use (i.e., equivalent feed). NRC staff is clarifying its current alternate feed policy to declare IX resins as equivalent feed. This clarification is necessary to alleviate a regulatory and financial burden on facilities that filter uranium using IX resin, such as CWSs and mine dewatering operations. Disposing of those resins in a licensed facility could be 40 to 50 percent of the total operations and maintenance (O and M) cost for a CWS. Allowing uranium recovery facilities to treat these resins without requiring a license amendment lowers O and M costs and captures a valuable natural resource. (authors)« less

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...

VISION User Guide - VISION (Verifiable Fuel Cycle Simulation) Model

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

Jacob J. Jacobson; Robert F. Jeffers; Gretchen E. Matthern

2009-08-01

The purpose of this document is to provide a guide for using the current version of the Verifiable Fuel Cycle Simulation (VISION) model. This is a complex model with many parameters; the user is strongly encouraged to read this user guide before attempting to run the model. This model is an R&D work in progress and may contain errors and omissions. It is based upon numerous assumptions. This model is intended to assist in evaluating “what if” scenarios and in comparing fuel, reactor, and fuel processing alternatives at a systems level for U.S. nuclear power. The model is not intendedmore » as a tool for process flow and design modeling of specific facilities nor for tracking individual units of fuel or other material through the system. The model is intended to examine the interactions among the components of a fuel system as a function of time varying system parameters; this model represents a dynamic rather than steady-state approximation of the nuclear fuel system. VISION models the nuclear cycle at the system level, not individual facilities, e.g., “reactor types” not individual reactors and “separation types” not individual separation plants. Natural uranium can be enriched, which produces enriched uranium, which goes into fuel fabrication, and depleted uranium (DU), which goes into storage. Fuel is transformed (transmuted) in reactors and then goes into a storage buffer. Used fuel can be pulled from storage into either separation of disposal. If sent to separations, fuel is transformed (partitioned) into fuel products, recovered uranium, and various categories of waste. Recycled material is stored until used by its assigned reactor type. Note that recovered uranium is itself often partitioned: some RU flows with recycled transuranic elements, some flows with wastes, and the rest is designated RU. RU comes out of storage if needed to correct the U/TRU ratio in new recycled fuel. Neither RU nor DU are designated as wastes. VISION is comprised of several Microsoft Excel input files, a Powersim Studio core, and several Microsoft Excel output files. All must be co-located in the same folder on a PC to function. We use Microsoft Excel 2003 and have not tested VISION with Microsoft Excel 2007. The VISION team uses both Powersim Studio 2005 and 2009 and it should work with either.« 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...

Data Mining Techniques to Estimate Plutonium, Initial Enrichment, Burnup, and Cooling Time in Spent Fuel Assemblies

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

Trellue, Holly Renee; Fugate, Michael Lynn; Tobin, Stephen Joesph

The Next Generation Safeguards Initiative (NGSI), Office of Nonproliferation and Arms Control (NPAC), National Nuclear Security Administration (NNSA) of the U.S. Department of Energy (DOE) has sponsored a multi-laboratory, university, international partner collaboration to (1) detect replaced or missing pins from spent fuel assemblies (SFA) to confirm item integrity and deter diversion, (2) determine plutonium mass and related plutonium and uranium fissile mass parameters in SFAs, and (3) verify initial enrichment (IE), burnup (BU), and cooling time (CT) of facility declaration for SFAs. A wide variety of nondestructive assay (NDA) techniques were researched to achieve these goals [Veal, 2010 andmore » Humphrey, 2012]. In addition, the project includes two related activities with facility-specific benefits: (1) determination of heat content and (2) determination of reactivity (multiplication). In this research, a subset of 11 integrated NDA techniques was researched using data mining solutions at Los Alamos National Laboratory (LANL) for their ability to achieve the above goals.« less

Neutronics and Transient Calculations for the Conversion of the Transient Reactor Rest Facility (TREAT)

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

Kontogeorgakos, Dimitrios C.; Connaway, Heather M.; Papadias, Dionissios D.

2015-01-01

The Transient Reactor Test Facility (TREAT) is a graphite-reflected, graphitemoderated, and air-cooled reactor fueled with 93.1% enriched UO2 particles dispersed in graphite, with a carbon-to-235U ratio of ~10000:1. TREAT was used to simulate accident conditions by subjecting fuel test samples placed at the center of the core to high energy transient pulses. The transient pulse production is based on the core’s selflimiting nature due to the negative reactivity feedback provided by the fuel graphite as the core temperature rises. The analysis of the conversion of TREAT to low enriched uranium (LEU) is currently underway. This paper presents the analytical methodsmore » used to calculate the transient performance of TREAT in terms of power pulse production and resulting peak core temperatures. The validation of the HEU neutronics TREAT model, the calculation of the temperature distribution and the temperature reactivity feedback as well as the number of fissions generated inside fuel test samples are discussed.« less

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.

Gamma-Ray Holdup Measurements of U-235, Np-237, and Am-241 Content in the C and D Out-gassing Ovens in the Deactivation and Decommissioning Activities in 321-M

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

RAYMOND, DEWBERRY

2004-09-16

The Analytical Development Section of Savannah River National Laboratory (SRNL) was requested by the Facilities Disposition Projects (FDP) 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 and Accountability, and to meet criticality safety controls. This report covers holdup measurements in the C and D out-gassing ovensmore » that were used to remove gas entrained in billet assembly material prior to the billets being extruded into rods by the extrusion press. A portable high purity germanium (HPGe) detection system and a portable sodium iodide (NaI) detection system were used to determine highly enriched uranium (HEU) holdup and to determine holdup of Np-237 and Am-241 that were observed in these components. The HPGe detector was run by an EG and G Dart (TM) system that contains the high voltage power supply and signal processing electronics. A personal computer with Gamma-Vision software was used to control the Dart (TM) MCA and provide space to store and manipulate multiple 4096-channel spectra. The NaI detector was run with a Canberra NaI plus MCA card that converts a personal computer to a full function multichannel analyzer and with Canberra Genie-2000 acquisition and analysis software. The measured Np-237 and Am-241 contents were especially important in these components because their presence is unusual and unexpected in 321-M. It was important to obtain a measured value of these two species to disposition the out-gassing ovens and to determine whether a separate waste stream was necessary for release of these contaminated components to the E-Area Solid Waste Vault. The reported values for Np-237 are (17 plus or minus 7) mg in oven C and less than 0.5 mg in oven D. The reported values for Am-241 are (1.3 plus or minus 0.2) in oven C and less than 400 ng in oven D. Our results indicate an upper limit of U-235 content of 0.2 g for oven C and (0.105 plus or minus 0.048) g in oven D. This report discusses the methodology, non-destructive assay (NDA) measurements, and results of the holdup measured for each of the three actinide species in these out-gassing ovens.« less

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

76 FR 72006 - Draft Interim Staff Guidance: Evaluations of Uranium Recovery Facility Surveys of Radon and Radon...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-11-21

.... Discussion Uranium recovery facility licensees, including in-situ recovery facilities and conventional... Recovery Facility Surveys of Radon and Radon Progeny in Air and Demonstrations of Compliance AGENCY... Staff Guidance, ``Evaluations of Uranium Recovery Facility Surveys of Radon and Radon Progeny in Air and...

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...

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.

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...

Occupational dose reduction at Department of Energy contractor facilities: Bibliography of selected readings in radiation protection and ALARA

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

Dionne, B.J.; Sullivan, S.G.; Baum, J.W.

1993-12-01

This bibliography contains abstracts relating to various aspects of ALARA program implementation and dose reduction activities, with a focus on DOE facilities. Abstracts included in this bibliography were selected from proceedings of technical meetings, journals, research reports, searches of the DOE Energy, Science and Technology Database (in general, the citation and abstract information is presented as obtained from this database), and reprints of published articles provided by the authors. Facility types and activities covered in the scope of this report include: radioactive waste, uranium enrichment, fuel fabrication, spent fuel storage and reprocessing, facility decommissioning, hot laboratories, tritium production, research, testmore » and production reactors, weapons fabrication and testing, fusion, uranium and plutonium processing, radiography, and aocelerators. Information on improved shielding design, decontamination, containments, robotics, source prevention and control, job planning, improved operational and design techniques, as well as on other topics, has been included. In addition, DOE/EH reports not included in previous volumes of the bibliography are in this volume (abstracts 611 to 684). This volume (Volume 5 of the series) contains 217 abstracts. An author index and a subject index are provided to facilitate use. Both indices contain the abstract numbers from previous volumes, as well as the current volume. Information that the reader feels might be included in the next volume of this bibliography should be submitted to the BNL ALARA Center.« less

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

Hughes, Joan F.

The US Department of Energy’s (DOE’s) Oak Ridge Reservation (ORR) is located in Roane and Anderson counties in East Tennessee, about 40 km (25 miles) from Knoxville. ORR is one of DOE’s most unique and complex sites. It encompasses three major facilities and thousands of employees that perform every mission in the DOE portfolio—energy research, environmental restoration, national security, nuclear fuel supply, reindustrialization, science education, basic and applied research in areas important to US security, and technology transfer. ORR was established in the early 1940s as part of the Manhattan Project for the purposes of enriching uranium and pioneering methodsmore » for producing and separating plutonium. Today, scientists at the Oak Ridge National Laboratory (ORNL), DOE’s largest multipurpose national laboratory, conduct world-leading research in advanced materials, alternative fuels, climate change, and supercomputing. The Y-12 National Security Complex (Y-12 or Y-12 Complex) is vital to maintaining the safety, security, and effectiveness of the US nuclear weapons stockpile and reducing the global threat posed by nuclear proliferation and terrorism. The East Tennessee Technology Park (ETTP), a former uranium enrichment complex, is being transitioned to a clean, revitalized industrial park.« less

An Optically Stimulated Luminescence Uranium Enrichment Monitor

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

Miller, Steven D.; Tanner, Jennifer E.; Simmons, Kevin L.

The Pacific Northwest National Laboratory (PNNL) has pioneered the use of Optically Stimulated Luminescence (OSL) technology for use in personnel dosimetry and high dose radiation processing dosimetry. PNNL has developed and patented an alumina-based OSL dosimeter that is being used by the majority of medical X-ray and imaging technicians worldwide. PNNL has conceived of using OSL technology to passively measure the level of UF6 enrichment by attaching the prototype OSL monitor to pipes containing UF6 gas within an enrichment facility. The prototype OSL UF6 monitor utilizes a two-element approach with the first element open and unfiltered to measure both themore » low energy and high energy gammas from the UF6, while the second element utilizes a 3-mm thick tungsten filter to eliminate the low energy gammas and pass only the high energy gammas from the UF6. By placing a control monitor in the room away from the UF6 pipes and other ionizing radiation sources, the control readings can be subtracted from the UF6 pipe monitor measurements. The ratio of the shielded to the unshielded net measurements provides a means to estimate the level of uranium enrichment. PNNL has replaced the commercially available MicroStar alumina-based dosimeter elements with a composite of polyethylene plastic, high-Z glass powder, and BaFBr:Eu OSL phosphor powder at various concentrations. The high-Z glass was added in an attempt to raise the average “Z” of the composite dosimeter and increase the response. Additionally, since BaFBr:Eu OSL phosphor is optimally excited and emits light at different wavelengths compared to alumina, the commercially available MicroStar reader was modified for reading BaFBr:Eu in a parallel effort to increase reader sensitivity. PNNL will present the design and performance of our novel OSL uranium enrichment monitor based on a combination of laboratory and UF6 test loop measurements. PNNL will also report on the optimization effort to achieve the highest possible performance from both the OSL enrichment monitor and the new custom OSL reader modified for this application. This project has been supported by the US Department of Energy’s National Nuclear Security Administration’s Office of Dismantlement and Transparency (DOE/NNSA/NA-241).« less

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

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...

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

Uranium Enrichment Safeguards

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

Demuth, Scott F.; Trahan, Alexis Chanel

2017-06-26

DIV of facility layout, material flows, and other information provided in the DIQ. Material accountancy through an annual PIV and a number of interim inventory verifications, including UF6 cylinder identification and counting, NDA of cylinders, and DA on a sample collection of UF6. Application of C/S technologies utilizing seals and tamper-indicating devices (TIDs) on cylinders, containers, storage rooms, and IAEA instrumentation to provide continuity of knowledge between inspection. Verification of the absence of undeclared material and operations, especially HEU production, through SNRIs, LFUA of cascade halls, and environmental swipe sampling

Secure Transportation of HEU in Romania

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

None

2009-07-06

The National Nuclear Security Administration has announced the final shipments of Russian-origin highly enriched uranium (HEU) nuclear fuel from Romania. The material was removed and returned to Russia by air for storage at two secure nuclear facilities, making Romania the first country to remove all HEU since President Obama outlined his commitment to securing all vulnerable nuclear material around the world within four years. This was also the first time NNSA has shipped spent HEU by airplane, a development that will help accelerate efforts to meet the Presidents objective.

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

Cozzi, A.; Johnson, F.

Production of Mo-99 for medical isotope use is being investigated using dissolved low enriched uranium (LEU) fissioned using an accelerator driven process. With the production and separation of Mo-99, a low level waste stream will be generated. Since the production facility is a commercial endeavor, waste disposition paths normally available for federally generated radioactive waste may not be available. Disposal sites for commercially generated low level waste are available, and consideration to the waste acceptance criteria (WAC) of the disposal site should be integral in flowsheet development for the Mo-99 production. Pending implementation of the “Uranium Lease and Take-Back Programmore » for Irradiation for Production of Molybdenum-99 for Medical Use” as directed by the American Medical Isotopes Production Act of 2012, there are limited options for disposing of the waste generated by the production of Mo-99 using an accelerator. The commission of a trade study to assist in the determination of the most favorable balance of production throughput and waste management should be undertaken. The use of a waste broker during initial operations of a facility has several benefits that can offset the cost associated with using a subcontractor. As the facility matures, the development of in-house capabilities can be expanded to incrementally reduce the dependence on a subcontractor.« 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...

User Guide for VISION 3.4.7 (Verifiable Fuel Cycle Simulation) Model

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

Jacob J. Jacobson; Robert F. Jeffers; Gretchen E. Matthern

2011-07-01

The purpose of this document is to provide a guide for using the current version of the Verifiable Fuel Cycle Simulation (VISION) model. This is a complex model with many parameters and options; the user is strongly encouraged to read this user guide before attempting to run the model. This model is an R&D work in progress and may contain errors and omissions. It is based upon numerous assumptions. This model is intended to assist in evaluating 'what if' scenarios and in comparing fuel, reactor, and fuel processing alternatives at a systems level. The model is not intended as amore » tool for process flow and design modeling of specific facilities nor for tracking individual units of fuel or other material through the system. The model is intended to examine the interactions among the components of a fuel system as a function of time varying system parameters; this model represents a dynamic rather than steady-state approximation of the nuclear fuel system. VISION models the nuclear cycle at the system level, not individual facilities, e.g., 'reactor types' not individual reactors and 'separation types' not individual separation plants. Natural uranium can be enriched, which produces enriched uranium, which goes into fuel fabrication, and depleted uranium (DU), which goes into storage. Fuel is transformed (transmuted) in reactors and then goes into a storage buffer. Used fuel can be pulled from storage into either separation or disposal. If sent to separations, fuel is transformed (partitioned) into fuel products, recovered uranium, and various categories of waste. Recycled material is stored until used by its assigned reactor type. VISION is comprised of several Microsoft Excel input files, a Powersim Studio core, and several Microsoft Excel output files. All must be co-located in the same folder on a PC to function. You must use Powersim Studio 8 or better. We have tested VISION with the Studio 8 Expert, Executive, and Education versions. The Expert and Education versions work with the number of reactor types of 3 or less. For more reactor types, the Executive version is currently required. The input files are Excel2003 format (xls). The output files are macro-enabled Excel2007 format (xlsm). VISION 3.4 was designed with more flexibility than previous versions, which were structured for only three reactor types - LWRs that can use only uranium oxide (UOX) fuel, LWRs that can use multiple fuel types (LWR MF), and fast reactors. One could not have, for example, two types of fast reactors concurrently. The new version allows 10 reactor types and any user-defined uranium-plutonium fuel is allowed. (Thorium-based fuels can be input but several features of the model would not work.) The user identifies (by year) the primary fuel to be used for each reactor type. The user can identify for each primary fuel a contingent fuel to use if the primary fuel is not available, e.g., a reactor designated as using mixed oxide fuel (MOX) would have UOX as the contingent fuel. Another example is that a fast reactor using recycled transuranic (TRU) material can be designated as either having or not having appropriately enriched uranium oxide as a contingent fuel. Because of the need to study evolution in recycling and separation strategies, the user can now select the recycling strategy and separation technology, by year.« less

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.

ANALYSIS OF 2H-EVAPORATOR SCALE WALL [HTF-13-82] AND POT BOTTOM [HTF-13-77] SAMPLES

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

Oji, L.

2013-06-21

Savannah River Remediation (SRR) is planning to remove a buildup of sodium aluminosilicate scale from the 2H-evaporator pot by loading and soaking the pot with heated 1.5 M nitric acid solution. Sampling and analysis of the scale material has been performed so that uranium and plutonium isotopic analysis can be input into a Nuclear Criticality Safety Assessment (NCSA) for scale removal by chemical cleaning. Historically, since the operation of the Defense Waste Processing Facility (DWPF), silicon in the DWPF recycle stream combines with aluminum in the typical tank farm supernate to form sodium aluminosilicate scale mineral deposits in the 2Hevaporatormore » pot and gravity drain line. The 2H-evaporator scale samples analyzed by Savannah River National Laboratory (SRNL) came from the bottom cone sections of the 2H-evaporator pot [Sample HTF-13-77] and the wall 2H-evaporator [sample HTF-13-82]. X-ray diffraction analysis (XRD) confirmed that both the 2H-evaporator pot scale and the wall samples consist of nitrated cancrinite (a crystalline sodium aluminosilicate solid) and clarkeite (a uranium oxy-hydroxide mineral). On “as received” basis, the bottom pot section scale sample contained an average of 2.59E+00 ± 1.40E-01 wt % total uranium with a U-235 enrichment of 6.12E-01 ± 1.48E-02 %, while the wall sample contained an average of 4.03E+00 ± 9.79E-01 wt % total uranium with a U-235 enrichment of 6.03E-01% ± 1.66E-02 wt %. The bottom pot section scale sample analyses results for Pu-238, Pu-239, and Pu-241 are 3.16E- 05 ± 5.40E-06 wt %, 3.28E-04 ± 1.45E-05 wt %, and <8.80E-07 wt %, respectively. The evaporator wall scale samples analysis values for Pu-238, Pu-239, and Pu-241 averages 3.74E-05 ± 6.01E-06 wt %, 4.38E-04 ± 5.08E-05 wt %, and <1.38E-06 wt %, respectively. The Pu-241 analyses results, as presented, are upper limit values. These results are provided so that SRR can calculate the equivalent uranium-235 concentrations for the NCSA. Results confirm that the uranium contained in the scale remains depleted with respect to natural uranium. SRNL did not calculate an equivalent U-235 enrichment, which takes into account other fissionable isotopes U-233, Pu-239 and Pu-241. The applicable method for calculation of equivalent U-235 will be determined in the NCSA.« less

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

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

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

Short, Steven M.; Phillips, Jon R.; Weimar, Mark R.

2008-09-01

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

Final Environmental Impact Statement for the construction and operation of Claiborne Enrichment Center, Homer, Louisiana (Docket No. 70-3-70). Volume 2, Public comments and NRC response

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

Zeitoun, A.

The Final Environmental Impact Statement (FEIS) (Volume 1), was prepared by the Nuclear Regulatory Commission (NRC) in accordance with regulation 10 CFR Part 51, which implements the National Environmental Policy Act (NEPA), to assess the potential environmental impacts for licensing the construction and operation of a proposed gaseous centrifuge enrichment facility to be built in Claiborne Parish, Louisiana by Louisiana Energy Services, L.P. (LES). The proposed facility would have a production capacity of about 866 metric tons annually of up to 5 weight percent enriched UF{sub 6}, using a proven centrifuge technology. Included in the assessment are co on, bothmore » normal operations and potential accidents (internal and external events), and the eventual decontamination and decommissioning of the site. In order to help assure that releases from the operation of the facility and potential impacts on the public are as low as reasonably achievable, an environmental monitoring program was developed by LES to detect significant changes in the background levels of uranium around the site. Other issues addressed include the purpose and need for the facility, the alternatives to the proposed action, potential disposition of the tails, the site selection process, and environmental justice. The NRC staff concludes that the facility can be constructed and operated with small and acceptable impacts on the public and the environment, and proposes to issue a license to the applicant, Louisiana Energy Services, to authorize construction and operation of the proposed facility. The letters in this Appendix have been divided into three sections. Section One contains letters to which the NRC responded by addressing specific comments. Section Two contains the letters that concerned the communities of Forest Grove and Center Springs. Section Three is composed of letters that required no response. These letters were generally in support of the facility.« less

Using the Time-Correlated Induced Fission Method to Simultaneously Measure the 235U Content and the Burnable Poison Content in LWR Fuel Assemblies

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

Root, M. A.; Menlove, H. O.; Lanza, R. C.

The uranium neutron coincidence collar uses thermal neutron interrogation to verify the 235U mass in low-enriched uranium (LEU) fuel assemblies in fuel fabrication facilities. Burnable poisons are commonly added to nuclear fuel to increase the lifetime of the fuel. The high thermal neutron absorption by these poisons reduces the active neutron signal produced by the fuel. Burnable poison correction factors or fast-mode runs with Cd liners can help compensate for this effect, but the correction factors rely on operator declarations of burnable poison content, and fast-mode runs are time-consuming. Finally, this paper describes a new analysis method to measure themore » 235U mass and burnable poison content in LEU nuclear fuel simultaneously in a timely manner, without requiring additional hardware.« less

Using the Time-Correlated Induced Fission Method to Simultaneously Measure the 235U Content and the Burnable Poison Content in LWR Fuel Assemblies

DOE PAGES

Root, M. A.; Menlove, H. O.; Lanza, R. C.; ...

2018-03-21

The uranium neutron coincidence collar uses thermal neutron interrogation to verify the 235U mass in low-enriched uranium (LEU) fuel assemblies in fuel fabrication facilities. Burnable poisons are commonly added to nuclear fuel to increase the lifetime of the fuel. The high thermal neutron absorption by these poisons reduces the active neutron signal produced by the fuel. Burnable poison correction factors or fast-mode runs with Cd liners can help compensate for this effect, but the correction factors rely on operator declarations of burnable poison content, and fast-mode runs are time-consuming. Finally, this paper describes a new analysis method to measure themore » 235U mass and burnable poison content in LEU nuclear fuel simultaneously in a timely manner, without requiring additional hardware.« 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...

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

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...

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.

Assay of Drums with Unknown Content Stored in 247-41F

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

Dewberry, R.

The Analytical Development Section of Savannah River Technology Center (SRTC) was requested by the Facilities Decontamination and Decommissioning Program (FDD) to determine the radionuclide content in two drums that were stored in an inactive warehouse of the Naval Fuels facility. The drums were labeled as containing fissile material and were placed in a critically safe arrangement, but it was not known whether they still contained the fissile material. Our g-PHA assay results indicate that the unknown highly enriched uranium (HEU) content of the two drums is one and 0.5 grams of surface contamination. Our neutron measurements confirmed that there aremore » no significant lumps of 235U present in these drums and that only surface contamination is present. The results confirmed that the facility was in compliance with administrative controls for fissile materials and that it is safe to open the drums for visual inspection.« less

Neutralization of Plutonium and Enriched Uranium Solutions Containing Gadolinium as a Neutron Poison

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

BRONIKOWSKI, MG.

2004-04-01

Materials currently being dissolved in the HB-Line Facility will result in an accumulated solution containing an estimated uranium:plutonium (U:Pu) ratio of 4.3:1 and an 235U enrichment estimated at 30 per cent The U:Pu ratio and the enrichment are outside the evaluated concentration range for disposition to high level waste (HLW) using gadolinium (Gd) as a neutron poison. To confirm that the solution generated during the current HB-Line dissolving campaign can be poisoned with Gd, neutralized and discarded to the Savannah River Site (SRS) high level waste (HLW) system without undue nuclear safety concerns the caustic precipitation of surrogate solutions wasmore » examined. Experiments were performed with a U/Pu/Gd solution representative of the HB-Line estimated concentration ratio and also a U/Gd solution. Depleted U was used in the experiments as the enrichment of the U will not affect the chemical behavior during neutralization, but will affect the amount of Gd added to the solution. Settling behavior of the neutralized solutions was found to be comparable to previous studies. The neutralized solutions mixed easily and had expected densities of typical neutralized waste. The neutralized solids were found to be homogeneous and less than 20 microns in size. Partially neutralized solids were more amorphous than the fully neutralized solids. Based on the results of these experiments, Gd was found to be a viable poison for neutralizing a U/Pu/Gd solution with a U:Pu mass ratio of 4.3:1 thus extending the U:Pu mass ratio from the previously investigated 0-3:1 to 4.3:1. However, further work is needed to allow higher U concentrations or U:Pu ratios greater than investigated in this work.« less

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

Overview and Current Status of Analyses of Potential LEU Design Concepts for TREAT

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

Connaway, H. M.; Kontogeorgakos, D. C.; Papadias, D. D.

2015-10-01

Neutronic and thermal-hydraulic analyses have been performed to evaluate the performance of different low-enriched uranium (LEU) fuel design concepts for the conversion of the Transient Reactor Test Facility (TREAT) from its current high-enriched uranium (HEU) fuel. TREAT is an experimental reactor developed to generate high neutron flux transients for the testing of nuclear fuels. The goal of this work was to identify an LEU design which can maintain the performance of the existing HEU core while continuing to operate safely. A wide variety of design options were considered, with a focus on minimizing peak fuel temperatures and optimizing the powermore » coupling between the TREAT core and test samples. Designs were also evaluated to ensure that they provide sufficient reactivity and shutdown margin for each control rod bank. Analyses were performed using the core loading and experiment configuration of historic M8 Power Calibration experiments (M8CAL). The Monte Carlo code MCNP was utilized for steady-state analyses, and transient calculations were performed with the point kinetics code TREKIN. Thermal analyses were performed with the COMSOL multi-physics code. Using the results of this study, a new LEU Baseline design concept is being established, which will be evaluated in detail in a future report.« less

The ``Nuclear Renaissance'' and the Spread of Nuclear Weapons

NASA Astrophysics Data System (ADS)

Lyman, Edwin S.

2007-05-01

As interest grows around the world in nuclear power as an energy source that could help control greenhouse gas emissions, some have proclaimed the arrival of a ``nuclear renaissance.'' But can the increased risks of more nuclear power be managed? The political crisis surrounding Iran's pursuit of uranium enrichment has exposed weaknesses in the nuclear nonproliferation regime. Also, al Qaeda's declared interest in weapons of mass destruction raises the concern that terrorists could acquire nuclear weapons by stealing materials from poorly secured facilities. Growth of nuclear energy would require the construction of many additional uranium enrichment plants. And the generation of more spent nuclear fuel without a credible waste disposal strategy would increase political support for reprocessing, which separates large quantities of weapon-usable plutonium from spent fuel. There is little evidence that the various institutional arrangements and technical schemes proposed to mitigate the security risks of a major nuclear expansion would be effective. This talk will focus on the measures necessary to allow large-scale global growth of nuclear power without resulting in an unacceptably high risk of nuclear proliferation and nuclear terrorism, and will discuss the feasibility of such measures. To cite this abstract, use the following reference: http://meetings.aps.org/link/BAPS.2007.OSS07.E1.2

Modeling and Depletion Simulations for a High Flux Isotope Reactor Cycle with a Representative Experiment Loading

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

Chandler, David; Betzler, Ben; Hirtz, Gregory John

2016-09-01

The purpose of this report is to document a high-fidelity VESTA/MCNP High Flux Isotope Reactor (HFIR) core model that features a new, representative experiment loading. This model, which represents the current, high-enriched uranium fuel core, will serve as a reference for low-enriched uranium conversion studies, safety-basis calculations, and other research activities. A new experiment loading model was developed to better represent current, typical experiment loadings, in comparison to the experiment loading included in the model for Cycle 400 (operated in 2004). The new experiment loading model for the flux trap target region includes full length 252Cf production targets, 75Se productionmore » capsules, 63Ni production capsules, a 188W production capsule, and various materials irradiation targets. Fully loaded 238Pu production targets are modeled in eleven vertical experiment facilities located in the beryllium reflector. Other changes compared to the Cycle 400 model are the high-fidelity modeling of the fuel element side plates and the material composition of the control elements. Results obtained from the depletion simulations with the new model are presented, with a focus on time-dependent isotopic composition of irradiated fuel and single cycle isotope production metrics.« less

Summary of ORSphere critical and reactor physics measurements

NASA Astrophysics Data System (ADS)

Marshall, Margaret A.; Bess, John D.

2017-09-01

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) 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 the GODIVA I experiments. This critical configuration has been evaluated. Preliminary results were presented at ND2013. Since then, the evaluation was finalized and judged to be an acceptable benchmark experiment for the International Criticality Safety Benchmark Experiment Project (ICSBEP). Additionally, reactor physics measurements were performed to determine surface button worths, central void worth, delayed neutron fraction, prompt neutron decay constant, fission density and neutron importance. These measurements have been evaluated and found to be acceptable experiments and are discussed in full detail in the International Handbook of Evaluated Reactor Physics Benchmark Experiments. The purpose of this paper is to summarize all the evaluated critical and reactor physics measurements evaluations.

Occupational dose reduction at Department of Energy contractor facilities: Bibliography of selected readings in radiation protection and ALARA; Volume 5

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

Dionne, B.J.; Sullivan, S.G.; Baum, J.W.

1994-01-01

Promoting the exchange of information related to implementation of the As Low as Reasonably Achievable (ALARA) philosophy is a continuing objective for the Department of Energy (DOE). This report was prepared by the Brookhaven National Laboratory (BNL) ALARA Center for the DOE Office of Health. It contains the fifth in a series of bibliographies on dose reduction at DOE facilities. The BNL ALARA Center was originally established in 1983 under the sponsorship of the Nuclear Regulatory Commission to monitor dose-reduction research and ALARA activities at nuclear power plants. This effort was expanded in 1988 by the DOE`s Office of Environment,more » Safety and Health, to include DOE nuclear facilities. This bibliography contains abstracts relating to various aspects of ALARA program implementation and dose-reduction activities, with a specific focus on DOE facilities. Abstracts included in this bibliography were selected from proceedings of technical meetings, journals, research reports, searches of the DOE Energy, Science and Technology Database (in general, the citation and abstract information is presented as obtained from this database), and reprints of published articles provided by the authors. Facility types and activities covered in the scope of this report include: radioactive waste, uranium enrichment, fuel fabrication, spent fuel storage and reprocessing, facility decommissioning, hot laboratories, tritium production, research, test and production reactors, weapons fabrication and testing, fusion, uranium and plutonium processing, radiography, and accelerators. Information on improved shielding design, decontamination, containments, robotics, source prevention and control, job planning, improved operational and design techniques, as well as on other topics, has been included. In addition, DOE/EH reports not included in previous volumes of the bibliography are in this volume (abstracts 611 to 684). This volume (Volume 5 of the series) contains 217 abstracts.« less

Interlaboratory comparison program for nondestructive assay of prototype uranium reference materials

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

Trahey, N.M.; Smith, M.M.; Voeks, A.M.

The US Department of Energy (DOE), New Brunswick Laboratory (NBS), designed and administered an interlaboratory comparison program based on the measurement of NBL-produced prototype uranium nondestructive assay (NDA) reference materials for scrap and waste. The objectives of the program were to evaluate the reliability of NDA techniques as applied to nuclear safeguards materials control and accountability needs and to investigate the feasibility of providing practical NDA scrap and waste reference materials for use throughout the nuclear safeguards community. Fourteen facilities representing seven DOE contractors, four US Nuclear Regulatory Commission (NRC) licensees, one EURATOM Laboratory, and NBL, participated in this program.more » Three stable, well-characterized uranium reference materials were developed and certified for this program. Synthetic calcined ash, cellulose fiber, and ion-exchange resin simulate selected uranium scrap and waste forms which are often encountered in fabrication and recovery operations. The synthetic calcined ash represents an intermediate density inorganic matrix while the cellulose fiber and ion-exchange resin are representative of low-density organic matrices. The materials, containing from 0 to 13% uranium enriched at 93% /sup 235/U, were sealed in specially selected containers. Nineteen prototype reference samples, plus three empty containers, one to accompany each set, was circulated to the participants between August 1979 and May 1984. Triplicate measurements for /sup 235/U on each of the 19 filled containers were required. In addition, participants could opt to perform modular configuration measurements using containers from Sets IIA and IIB to simulate non-homogeneously dispersed uranium in waste containers. All data were reported to NBL for evaluation.« less

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

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.

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

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...

40 CFR 421.320 - Applicability: Description of the secondary uranium subcategory.

Code of Federal Regulations, 2012 CFR

2012-07-01

... secondary uranium subcategory. 421.320 Section 421.320 Protection of Environment ENVIRONMENTAL PROTECTION... CATEGORY Secondary Uranium Subcategory § 421.320 Applicability: Description of the secondary uranium... uranium (including depleted uranium) by secondary uranium facilities. ...

40 CFR 421.320 - Applicability: Description of the secondary uranium subcategory.

Code of Federal Regulations, 2011 CFR

2011-07-01

... secondary uranium subcategory. 421.320 Section 421.320 Protection of Environment ENVIRONMENTAL PROTECTION... CATEGORY Secondary Uranium Subcategory § 421.320 Applicability: Description of the secondary uranium... uranium (including depleted uranium) by secondary uranium facilities. ...

40 CFR 421.320 - Applicability: Description of the secondary uranium subcategory.

Code of Federal Regulations, 2013 CFR

2013-07-01

... secondary uranium subcategory. 421.320 Section 421.320 Protection of Environment ENVIRONMENTAL PROTECTION... CATEGORY Secondary Uranium Subcategory § 421.320 Applicability: Description of the secondary uranium... uranium (including depleted uranium) by secondary uranium facilities. ...

40 CFR 421.320 - Applicability: Description of the secondary uranium subcategory.

Code of Federal Regulations, 2014 CFR

2014-07-01

... secondary uranium subcategory. 421.320 Section 421.320 Protection of Environment ENVIRONMENTAL PROTECTION... CATEGORY Secondary Uranium Subcategory § 421.320 Applicability: Description of the secondary uranium... uranium (including depleted uranium) by secondary uranium facilities. ...

40 CFR 421.320 - Applicability: Description of the secondary uranium subcategory.

Code of Federal Regulations, 2010 CFR

2010-07-01

... secondary uranium subcategory. 421.320 Section 421.320 Protection of Environment ENVIRONMENTAL PROTECTION... CATEGORY Secondary Uranium Subcategory § 421.320 Applicability: Description of the secondary uranium... uranium (including depleted uranium) by secondary uranium facilities. ...

The U.S. Geological Survey's TRIGA® reactor

USGS Publications Warehouse

DeBey, Timothy M.; Roy, Brycen R.; Brady, Sally R.

2012-01-01

The U.S. Geological Survey (USGS) operates a low-enriched uranium-fueled, pool-type reactor located at the Federal Center in Denver, Colorado. The mission of the Geological Survey TRIGA® Reactor (GSTR) is to support USGS science by providing information on geologic, plant, and animal specimens to advance methods and techniques unique to nuclear reactors. The reactor facility is supported by programs across the USGS and is organizationally under the Associate Director for Energy and Minerals, and Environmental Health. The GSTR is the only facility in the United States capable of performing automated delayed neutron analyses for detecting fissile and fissionable isotopes. Samples from around the world are submitted to the USGS for analysis using the reactor facility. Qualitative and quantitative elemental analyses, spatial elemental analyses, and geochronology are performed. Few research reactor facilities in the United States are equipped to handle the large number of samples processed at the GSTR. Historically, more than 450,000 sample irradiations have been performed at the USGS facility. Providing impartial scientific information to resource managers, planners, and other interested parties throughout the world is an integral part of the research effort of the USGS.

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

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

Christopher Landers; Igor Bolshinsky; Ken Allen

2010-07-01

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

Analysis of the TREAT LEU Conceptual Design

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

Connaway, H. M.; Kontogeorgakos, D. C.; Papadias, D. D.

2016-03-01

Analyses were performed to evaluate the performance of the low enriched uranium (LEU) conceptual design fuel for the conversion of the Transient Reactor Test Facility (TREAT) from its current highly enriched uranium (HEU) fuel. TREAT is an experimental nuclear reactor designed to produce high neutron flux transients for the testing of reactor fuels and other materials. TREAT is currently in non-operational standby, but is being restarted under the U.S. Department of Energy’s Resumption of Transient Testing Program. The conversion of TREAT is being pursued in keeping with the mission of the Department of Energy National Nuclear Security Administration’s Material Managementmore » and Minimization (M3) Reactor Conversion Program. The focus of this study was to demonstrate that the converted LEU core is capable of maintaining the performance of the existing HEU core, while continuing to operate safely. Neutronic and thermal hydraulic simulations have been performed to evaluate the performance of the LEU conceptual-design core under both steady-state and transient conditions, for both normal operation and reactivity insertion accident scenarios. In addition, ancillary safety analyses which were performed for previous LEU design concepts have been reviewed and updated as-needed, in order to evaluate if the converted LEU core will function safely with all existing facility systems. Simulations were also performed to evaluate the detailed behavior of the UO 2-graphite fuel, to support future fuel manufacturing decisions regarding particle size specifications. The results of these analyses will be used in conjunction with work being performed at Idaho National Laboratory and Los Alamos National Laboratory, in order to develop the Conceptual Design Report project deliverable.« less

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

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

Lell, R. M.; McKnight, R. D.; Tsiboulia, A.

2010-09-30

Over a period of 30 years, more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited for nuclear data validation and to form the basis for criticality safety benchmarks. A number of the Argonne ZPR/ZPPR critical assemblies have been evaluated as ICSBEP and IRPhEP benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physicsmore » benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. ZPR-3 Assembly 11 (ZPR-3/11) was designed as a fast reactor physics benchmark experiment with an average core {sup 235}U enrichment of approximately 12 at.% and a depleted uranium reflector. Approximately 79.7% of the total fissions in this assembly occur above 100 keV, approximately 20.3% occur below 100 keV, and essentially none below 0.625 eV - thus the classification as a 'fast' assembly. This assembly is Fast Reactor Benchmark No. 8 in the Cross Section Evaluation Working Group (CSEWG) Benchmark Specificationsa and has historically been used as a data validation benchmark assembly. Loading of ZPR-3 Assembly 11 began in early January 1958, and the Assembly 11 program ended in late January 1958. The core consisted of highly enriched uranium (HEU) plates and depleted uranium plates loaded into stainless steel drawers, which were inserted into the central square stainless steel tubes of a 31 x 31 matrix on a split table machine. The core unit cell consisted of two columns of 0.125 in.-wide (3.175 mm) HEU plates, six columns of 0.125 in.-wide (3.175 mm) depleted uranium plates and one column of 1.0 in.-wide (25.4 mm) depleted uranium plates. The length of each column was 10 in. (254.0 mm) in each half of the core. The axial blanket consisted of 12 in. (304.8 mm) of depleted uranium behind the core. The thickness of the depleted uranium radial blanket was approximately 14 in. (355.6 mm), and the length of the radial blanket in each half of the matrix was 22 in. (558.8 mm). The assembly geometry approximated a right circular cylinder as closely as the square matrix tubes allowed. According to the logbook and loading records for ZPR-3/11, the reference critical configuration was loading 10 which was critical on January 21, 1958. Subsequent loadings were very similar but less clean for criticality because there were modifications made to accommodate reactor physics measurements other than criticality. Accordingly, ZPR-3/11 loading 10 was selected as the only configuration for this benchmark. As documented below, it was determined to be acceptable as a criticality safety benchmark experiment. A very accurate transformation to a simplified model is needed to make any ZPR assembly a practical criticality-safety benchmark. There is simply too much geometric detail in an exact (as-built) model of a ZPR assembly, even a clean core such as ZPR-3/11 loading 10. The transformation must reduce the detail to a practical level without masking any of the important features of the critical experiment. And it must do this without increasing the total uncertainty far beyond that of the original experiment. Such a transformation is described in Section 3. It was obtained using a pair of continuous-energy Monte Carlo calculations. First, the critical configuration was modeled in full detail - every plate, drawer, matrix tube, and air gap was modeled explicitly. Then the regionwise compositions and volumes from the detailed as-built model were used to construct a homogeneous, two-dimensional (RZ) model of ZPR-3/11 that conserved the mass of each nuclide and volume of each region. The simple cylindrical model is the criticality-safety benchmark model. The difference in the calculated k{sub eff} values between the as-built three-dimensional model and the homogeneous two-dimensional benchmark model was used to adjust the measured excess reactivity of ZPR-3/11 loading 10 to obtain the k{sub eff} for the benchmark model.« less

Design of a 25-kWe Surface Reactor System Based on SNAP Reactor Technologies

NASA Astrophysics Data System (ADS)

Dixon, David D.; Hiatt, Matthew T.; Poston, David I.; Kapernick, Richard J.

2006-01-01

A Hastelloy-X clad, sodium-potassium (NaK-78) cooled, moderated spectrum reactor using uranium zirconium hydride (UZrH) fuel based on the SNAP program reactors is a promising design for use in surface power systems. This paper presents a 98 kWth reactor for a power system the uses multiple Stirling engines to produce 25 kWe-net for 5 years. The design utilizes a pin type geometry containing UZrHx fuel clad with Hastelloy-X and NaK-78 flowing around the pins as coolant. A compelling feature of this design is its use of 49.9% enriched U, allowing it to be classified as a category III-D attractiveness and reducing facility costs relative to highly-enriched space reactor concepts. Presented below are both the design and an analysis of this reactor's criticality under various safety and operations scenarios.

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

Collaborative Russian-US work in nuclear material protection, control and accounting at the Institute of Physics and Power Engineering. 2: Extension to additional facilities

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

Kuzin, V.V.; Pshakin, G.M.; Belov, A.P.

1996-12-31

During 1995, collaborative Russian-US nuclear material protection, control, and accounting (MPC and A) tasks at the Institute of Physics and Power Engineering (IPPE) in Obninsk, Russia focused on improving the protection of nuclear materials at the BFS Fast Critical Facility. BFS has tens of thousands of fuel disks containing highly enriched uranium and weapons-grade plutonium that are used to simulate the core configurations of experimental reactors in two critical assemblies. Completed tasks culminated in demonstrations of newly implemented equipment (Russian and US) and methods that enhanced the MPC and A at BFS through computerized accounting, nondestructive inventory verification measurements, personnelmore » identification and access control, physical inventory taking, physical protection, and video surveillance. The collaborative work with US Department of Energy national laboratories is now being extended. In 1996 additional tasks to improve MPC and A have been implemented at BFS, the Technological Laboratory for Fuel Fabrication (TLFF) the Central Storage Facility (CSF), and for the entire site. The TLFF reclads BFS uranium metal fuel disks (process operations and transfers of fissile material). The CSF contains many different types of nuclear material. MPC and A at these additional facilities will be integrated with that at BFS as a prototype site-wide approach. Additional site-wide tasks encompass communications and tamper-indicating devices. Finally, new storage alternatives are being implemented that will consolidate the more attractive nuclear materials in a better-protected nuclear island. The work this year represents not just the addition of new facilities and the site-wide approach, but the systematization of the MPC and A elements that are being implemented as a first step and the more comprehensive ones planned.« less

Nondestructive Assay Data Integration with the SKB-50 Assemblies - FY16 Update

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

Tobin, Stephen Joseph; Fugate, Michael Lynn; Trellue, Holly Renee

2016-10-28

A project to research the application of non-destructive assay (NDA) techniques for spent fuel assemblies is underway at the Central Interim Storage Facility for Spent Nuclear Fuel (for which the Swedish acronym is Clab) in Oskarshamn, Sweden. The research goals of this project contain both safeguards and non-safeguards interests. These nondestructive assay (NDA) technologies are designed to strengthen the technical toolkit of safeguard inspectors and others to determine the following technical goals more accurately; Verify initial enrichment, burnup, and cooling time of facility declaration for spent fuel assemblies; Detect replaced or missing pins from a given spent fuel assembly tomore » confirm its integrity; and Estimate plutonium mass and related plutonium and uranium fissile mass parameters in spent fuel assemblies. Estimate heat content, and measure reactivity (multiplication).« 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.

Safeguards on uranium ore concentrate? the impact of modern mining and milling process

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

Francis, Stephen

2013-07-01

Increased purity in uranium ore concentrate not only raises the question as to whether Safeguards should be applied to the entirety of uranium conversion facilities, but also as to whether some degree of coverage should be moved back to uranium ore concentrate production at uranium mining and milling facilities. This paper looks at uranium ore concentrate production across the globe and explores the extent to which increased purity is evident and the underlying reasons. Potential issues this increase in purity raises for IAEA's strategy on the Starting Point of Safeguards are also discussed.

76 FR 58049 - Atomic Safety and Licensing Board; Honeywell International, Inc.; Metropolis Works Uranium...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-09-19

... NUCLEAR REGULATORY COMMISSION [Docket No. 40-3392-MLA; ASLBP No. 11-910-01-MLA-BD01] Atomic Safety and Licensing Board; Honeywell International, Inc.; Metropolis Works Uranium Conversion Facility... assurance for its Metropolis Works uranium conversion facility in Metropolis, Illinois. \\1\\ LBP-11-19, 74...

Methodology and Software for Gross Defect Detection of Spent Nuclear Fuel at the Atucha-I Reactor [Novel Methodology and Software for Spent Fuel Gross Defect Detection at the Atucha-I Reactor

DOE PAGES

Sitaraman, Shivakumar; Ham, Young S.; Gharibyan, Narek; ...

2017-03-27

Here, fuel assemblies in the spent fuel pool are stored by suspending them in two vertically stacked layers at the Atucha Unit 1 nuclear power plant (Atucha-I). This introduces the unique problem of verifying the presence of fuel in either layer without physically moving the fuel assemblies. Given that the facility uses both natural uranium and slightly enriched uranium at 0.85 wt% 235U and has been in operation since 1974, a wide range of burnups and cooling times can exist in any given pool. A gross defect detection tool, the spent fuel neutron counter (SFNC), has been used at themore » site to verify the presence of fuel up to burnups of 8000 MWd/t. At higher discharge burnups, the existing signal processing software of the tool was found to fail due to nonlinearity of the source term with burnup.« less

Methodology and Software for Gross Defect Detection of Spent Nuclear Fuel at the Atucha-I Reactor [Novel Methodology and Software for Spent Fuel Gross Defect Detection at the Atucha-I Reactor

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

Sitaraman, Shivakumar; Ham, Young S.; Gharibyan, Narek

Here, fuel assemblies in the spent fuel pool are stored by suspending them in two vertically stacked layers at the Atucha Unit 1 nuclear power plant (Atucha-I). This introduces the unique problem of verifying the presence of fuel in either layer without physically moving the fuel assemblies. Given that the facility uses both natural uranium and slightly enriched uranium at 0.85 wt% 235U and has been in operation since 1974, a wide range of burnups and cooling times can exist in any given pool. A gross defect detection tool, the spent fuel neutron counter (SFNC), has been used at themore » site to verify the presence of fuel up to burnups of 8000 MWd/t. At higher discharge burnups, the existing signal processing software of the tool was found to fail due to nonlinearity of the source term with burnup.« less

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

Thermionic System Evaluation Test: Ya-21U System Topaz International Program

DTIC Science & Technology

1996-07-01

by enriched uranium dioxide (U02) fuel pellets, as illustrated by Figure 5. The work section of the converter contained 34 TFEs that provided power...power system. This feature permitted transportation of the highly enriched uranium oxide fuel in separate containers from the space power system and...by Figure 8. The radial reflector contained three safety and nine control drums. Each drum contained a section of boron carbide (B4C) neutron poison

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)

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...

NECTAR—A fission neutron radiography and tomography facility

NASA Astrophysics Data System (ADS)

Bücherl, T.; Lierse von Gostomski, Ch.; Breitkreutz, H.; Jungwirth, M.; Wagner, F. M.

2011-09-01

NECTAR (Neutron Computerized Tomography and Radiography) is a versatile facility for radiographic and tomographic investigations as well as for neutron activation experiments using fission neutrons. The radiation sources for this facility are two plates of highly enriched uranium situated in the moderator vessel in FRM II. Thermal neutrons originating from the main fuel element of the reactor generate in these plates fast neutrons. These can escape through a horizontal beam tube without moderation. The beam can be filtered and manipulated in order to reduce the accompanying gamma radiation and to match the specific experimental tasks. A summary of the main parameters required for experimental set-up and (quantitative) data evaluation is presented. The (measured) spectra of the neutron and gamma radiations are shown along with the effect of different filters on their behavior. The neutron and gamma fluxes, dose rates, L/ D-ratios, etc. and the main parameters of the actually used detection systems for neutron imaging are given, too.

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

Summary of ORSphere Critical and Reactor Physics Measurements

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

Marshall, Margaret A.; Bess, John D.

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) 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 the GODIVAmore » I experiments. This critical configuration has been evaluated. Preliminary results were presented at ND2013. Since then, the evaluation was finalized and judged to be an acceptable benchmark experiment for the International Criticality Safety Benchmark Experiment Project (ICSBEP). Additionally, reactor physics measurements were performed to determine surface button worths, central void worth, delayed neutron fraction, prompt neutron decay constant, fission density and neutron importance. These measurements have been evaluated and found to be acceptable experiments and are discussed in full detail in the International Handbook of Evaluated Reactor Physics Benchmark Experiments. The purpose of this paper is summary summarize all the critical and reactor physics measurements evaluations and, when possible, to compare them to GODIVA experiment results.« less

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

Margaret A. Marshall

In the early 1970’s 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 attempt to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950’s (HEU-MET-FAST-001). 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 withmore » the GODIVA I experiments. “The very accurate description of this sphere, as assembled, establishes it as an ideal benchmark for calculational methods and cross-section data files.” (Reference 1) While performing the ORSphere experiments care was taken to accurately document component dimensions (±0. 0001 in. for non-spherical parts), masses (±0.01 g), and material data The experiment was also set up to minimize the amount of structural material in the sphere proximity. A three part sphere was initially assembled with an average radius of 3.4665 in. and was then machined down to an average radius of 3.4420 in. (3.4425 in. nominal). These two spherical configurations were evaluated and judged to be acceptable benchmark experiments; however, the two experiments are highly correlated.« less

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.

76 FR 60941 - Policy Regarding Submittal of Amendments for Processing of Equivalent Feed at Licensed Uranium...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-09-30

... State-licensed uranium recovery site, either conventional, heap leach, or in situ recovery. DATES... types of new uranium recovery facilities (conventional mills, heap leach facilities, and in situ... from the ground for processing at a mill. Rather, the ore is processed in-situ with the resulting...

Studies in support of an SNM cutoff agreement: The PUREX exercise

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

Stanbro, W.D.; Libby, R.; Segal, J.

1995-07-01

On September 23, 1993, President Clinton, in a speech before the United Nations General Assembly, called for an international agreement banning the production of plutonium and highly enriched uranium for nuclear explosive purposes. A major element of any verification regime for such an agreement would probably involve inspections of reprocessing plants in Nuclear Nonproliferation Treaty weapons states. Many of these are large facilities built in the 1950s with no thought that they would be subject to international inspection. To learn about some of the problems that might be involved in the inspection of such large, old facilities, the Department ofmore » Energy, Office of Arms Control and Nonproliferation, sponsored a mock inspection exercise at the PUREX plant on the Hanford Site. This exercise examined a series of alternatives for inspections of the PUREX as a model for this type of facility at other locations. A series of conclusions were developed that can be used to guide the development of verification regimes for a cutoff agreement at reprocessing facilities.« less

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

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...

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

Hecker, Siegfried S.

Actions of the Government of the Democratic People's Republic of Korea have precipitated two nuclear crises in the past 10 years. The 1994 crisis was resolved through the 'Agreed Framework.' North Korea agreed to 'freeze' and eventually dismantle its nuclear program (with U.S. help to store spent fuel safely and under IAEA inspection). In return, the United States agreed (with the KEDO international consortium) to build two light-water reactors and supply North Korea with heavy-fuel oil until the reactors come on line. In addition, both sides agreed to move towards full normalization of relations, work for peace and security onmore » a nuclear-free Korean Peninsula, and work on strengthening the international nonproliferation regime. The second nuclear crisis erupted when North Korean Government officials allegedly admitted to having a clandestine uranium enrichment program when confronted with this accusation by U.S. officials in October 2002. The United States (through KEDO) suspended heavy-fuel oil shipments and North Korea responded by expelling the IAEA inspectors, withdrawing from the Nuclear Nonproliferation Treaty, and restarting its nuclear program in January 2003. The North Korean Government has invited Professor John Lewis of Stanford University, a China and North Korea scholar, for Track I1 discussions of nuclear and other key issues since 1987. In August 2003, Professor Lewis visited North Korea just before the first six-party talks, which were designed by the United States to solve the current nuclear crisis. Professor Lewis was invited back for the January 2004 visit. He asked Jack Pritchard, former U.S. special envoy for DRPK negotiations, and me to accompany him. Two Asian affairs staff specialists from the U.S. Senate Foreign Relations Committee also joined us. I will report on the visit to the Yongbyon Nuclear Scientific Research Center on January 8,2004. We toured the 5 MWe reactor, the 50 MWe reactor construction site, the spent fuel pool storage building, and the radiochemical laboratory. We concluded that North Korea has restarted its 5 MWe reactor (which produces roughly 6 kg of plutonium annually), it removed the 8000 spent fuel rods that were previously stored under IAEA safeguards from the spent fuel pool, and that it most likely extracted the 25 to 30 kg of plutonium contained in these fuel rods. Although North Korean officials showed us what they claimed was their plutonium metal product from this reprocessing campaign, we were not able to conclude definitively that it was in fact plutonium metal and that it came from the most recent reprocessing campaign. Nevertheless, our North Korean hosts demonstrated that they had the capability, the facility and requisite capacity, and the technical expertise to produce plutonium metal. We were not shown any facilities or had the opportunity to talk to technical or military experts who were able to address the issue of whether or not North Korea had a 'deterrent' as claimed - that is, we were not able to conclude that North Korea can build a nuclear device and that it can integrate nuclear devices into suitable delivery systems. On the matter of uranium enrichment programs, Vice Minister Kim Gye Gwan categorically denied that North Korea has a uranium enrichment program - he said, 'we have no program, no equipment, and no technical expertise for uranium enrichment.' Upon return to the United States, I shared my observations and analysis with U.S. Government officials in Washington, DC, including congressional testimony to the Senate Foreign Relations Committee and briefings to two House of Representative Committees.« less

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

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

LEACHING OF URANIUM ORES USING ALKALINE CARBONATES AND BICARBONATES AT ATMOSPHERIC PRESSURE

DOEpatents

Thunaes, A.; Brown, E.A.; Rabbits, A.T.; Simard, R.; Herbst, H.J.

1961-07-18

A method of leaching uranium ores containing sulfides is described. The method consists of adding a leach solution containing alkaline carbonate and alkaline bicarbonate to the ore to form a slurry, passing the slurry through a series of agitators, passing an oxygen containing gas through the slurry in the last agitator in the series, passing the same gas enriched with carbon dioxide formed by the decomposition of bicarbonates in the slurry through the penultimate agitator and in the same manner passing the same gas increasingly enriched with carbon dioxide through the other agitators in the series. The conditions of agitation is such that the extraction of the uranium content will be substantially complete before the slurry reaches the last agitator.

Development and Attestation of Gamma-Ray Measurement Methodologies for use by Rostekhnadzor Inspectors in the Russian Federation

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

Jeff Sanders

2006-09-01

Development and attestation of gamma-ray non-destructive assay measurement methodologies for use by inspectors of the Russian Federal Service for Environmental, Technological, and Nuclear Oversight (Rostekhnadzor, formerly Gosatomnadzor or GAN), as well as for use by Russian nuclear facilities, has been completed. Specifically, a methodology utilizing the gamma-ray multi group analysis (MGA) method for determining plutonium isotopic composition has been developed, while existing methodologies to determining uranium enrichment and isotopic composition have been revised to make them more appropriate to the material types and conditions present in nuclear facilities in the Russian Federation. This paper will discuss the development and revisionmore » of these methodologies, the metrological characteristics of the final methodologies, as well as the limitations and concerns specific to the utilization of these analysis methods in the Russian Federation.« less

The North Korean nuclear dilemma.

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

Hecker, Siegfried S.

2004-01-01

The current nuclear crisis, the second one in ten years, erupted when North Korea expelled international nuclear inspectors in December 2002, then withdrew from the Nuclear Nonproliferation Treaty (NPT), and claimed to be building more nuclear weapons with the plutonium extracted from the spent fuel rods heretofore stored under international inspection. These actions were triggered by a disagreement over U.S. assertions that North Korea had violated the Agreed Framework (which froze the plutonium path to nuclear weapons to end the first crisis in 1994) by clandestinely developing uranium enrichment capabilities providing an alternative path to nuclear weapons. With Stanford Universitymore » Professor John Lewis and three other Americans, I was allowed to visit the Yongbyon Nuclear Center on Jan. 8, 2004. We toured the 5 MWe reactor, the 50 MWe reactor construction site, the spent fuel pool storage building, and the radiochemical laboratory. We concluded that North Korea has restarted its 5 MWe reactor (which produces roughly 6 kg of plutonium annually), it removed the 8000 spent fuel rods that were previously stored under IAEA safeguards from the spent fuel pool, and that it most likely extracted the 25 to 30 kg of plutonium contained in these fuel rods. Although North Korean officials showed us what they claimed was their plutonium metal product from this reprocessing campaign, we were not able to conclude definitively that it was in fact plutonium metal and that it came from the most recent reprocessing campaign. Nevertheless, our North Korean hosts demonstrated that they had the capability, the facility and requisite capacity, and the technical expertise to produce plutonium metal. On the basis of our visit, we were not able to address the issue of whether or not North Korea had a 'deterrent' as claimed - that is, we were not able to conclude that North Korea can build a nuclear device and that it can integrate nuclear devices into suitable delivery systems. However, based on the capabilities we saw, we must assume that North Korea has the capability to produce a crude nuclear device. On the matter of uranium enrichment programs, our host categorically denied that North Korea has a uranium enrichment program - he said, 'we have no program, no equipment, and no technical expertise for uranium enrichment.' The denials were not convincing at the time and since then have proven to be quite hollow by the revelations of A.Q. Khan's nuclear black market activities. There is no easy solution to the nuclear crisis in North Korea. A military strike to eliminate the nuclear facilities was never very attractive and now has been overcome by events. The principal threat is posed by a stockpile of nuclear weapons and weapons-grade plutonium. We have no way of finding where either may be hidden. A diplomatic solution remains the only path forward, but it has proven elusive. All sides have proclaimed a nuclear weapons-free Korean Peninsula as the end goal. The U.S. Government has chosen to negotiate with North Korea by means of the six-party talks. It has very clearly outlined its position of insisting on complete, verifiable, irreversible dismantlement of all North Korean nuclear programs. North Korea has offered several versions of 're-freezing' its plutonium program while still denying a uranium enrichment program. It has insisted on simultaneous and reciprocal steps to a final solution. Regardless of which diplomatic path is chosen, the scientific challenges of eliminating the North Korean nuclear weapons programs (and its associated infrastructure) in a safe, secure, and verifiable manner are immense. The North Korean program is considerably more complex and developed than the fledgling Iraqi program of 1991 and Libyan program of 2004. It is more along the lines, but more complex than that of South Africa in the early 1990s. Actions taken or not taken by the North Koreans at their nuclear facilities during the course of the ongoing diplomatic discussions are key to whether or not the nuclear program can be eliminated safely and securely, and they will greatly influence the price tag for such operations. Moreover, they will determine whether or not one can verify complete elimination. Hence, cooperation of the North Koreans now and during the dismantlement and elimination stages is crucial. Technical discussions among specialists, perhaps within the framework of the working groups of the six-party talks, could be very productive in setting the stage for an effective, verifiable elimination of North Korea's nuclear weapons program.« less

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

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

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

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

Balanced program plan. Analysis for biomedical and environmental research

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

Not Available

1976-06-01

Major issues associated with the use of nuclear power are health hazards of exposure to radioactive materials; sources of radiation exposure; reactor accidents; sabotage of nuclear facilities; diversion of fissile material and its use for extortion; and the presence of plutonium in the environment. Fission fuel cycle technology is discussed with regard to milling, UF/sub 6/ production, uranium enrichment, plutonium fuel fabrication, power production, fuel processing, waste management, and fuel and waste transportation. The following problem areas of fuel cycle technology are briefly discussed: characterization, measurement, and monitoring; transport processes; health effects; ecological processes and effects; and integrated assessment. Estimatedmore » program unit costs are summarized by King-Muir Category. (HLW)« 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.

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.

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

Youker, Amanda J.; Krebs, John F.; Quigley, Kevin J.

With funding from the National Nuclear Security Administrations Material Management and Minimization Office, Argonne National Laboratory (Argonne) is providing technical assistance to help accelerate the U.S. production of Mo-99 using a non-highly enriched uranium (non-HEU) source. A potential Mo-99 production pathway is by accelerator-initiated fissioning in a subcritical uranyl sulfate solution containing low enriched uranium (LEU). As part of the Argonne development effort, we are undertaking the AMORE (Argonne Molybdenum Research Experiment) project, which is essentially a pilot facility for all phases of Mo-99 production, recovery, and purification. Production of Mo-99 and other fission products in the subcritical target solutionmore » is initiated by putting an electron beam on a depleted uranium (DU) target; the fast neutrons produced in the DU target are thermalized and lead to fissioning of U-235. At the end of irradiation, Mo is recovered from the target solution and separated from uranium and most of the fission products by using a titania column. The Mo is stripped from the column with an alkaline solution. After acidification of the Mo product solution from the recovery column, the Mo is concentrated (and further purified) in a second titania column. The strip solution from the concentration column is then purified with the LEU Modified Cintichem process. A full description of the process can be found elsewhere [1–3]. The initial commissioning steps for the AMORE project include performing a Mo-99 spike test with pH 1 sulfuric acid in the target vessel without a beam on the target to demonstrate the initial Mo separation-and-recovery process, followed by the concentration column process. All glovebox operations were tested with cold solutions prior to performing the Mo-99 spike tests. Two Mo-99 spike tests with pH 1 sulfuric acid have been performed to date. Figure 1 shows the flow diagram for the remotely operated Mo-recovery system for the AMORE project. There are two separate pumps and flow paths for the acid and base operations. The system contains three sample ladders with eight sample loops per ladder for target mixing; column loading, including acid and water washes; and column stripping, including the final water wash.« less

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.)

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

Performance evaluation of PRIDE UNDA system with pyroprocessing feed material.

PubMed

An, Su Jung; Seo, Hee; Lee, Chaehun; Ahn, Seong-Kyu; Park, Se-Hwan; Ku, Jeong-Hoe

2017-04-01

The PRIDE (PyRoprocessing Integrated inactive DEmonstration) is an engineering-scale pyroprocessing test-bed facility that utilizes depleted uranium (DU) instead of spent fuel as a process material. As part of the ongoing effort to enhance pyroprocessing safeguardability, UNDA (Unified Non-Destructive Assay), a system integrating three different non-destructive assay techniques, namely, neutron, gamma-ray, and mass measurement, for nuclear material accountancy (NMA) was developed. In the present study, UNDA's NMA capability was evaluated by measurement of the weight, 238 U mass, and U enrichment of oxide-reduction-process feed material (i.e., porous pellets). In the 238 U mass determination, the total neutron counts for porous pellets of six different weights were measured. The U enrichment of the porous pellets, meanwhile, was determined according to the gamma spectrums acquired using UNDA's NaI-based enrichment measurement system. The results demonstrated that the UNDA system, after appropriate corrections, could be used in PRIDE NMA applications with reasonable uncertainty. It is expected that in the near future, the UNDA system will be tested with next-step materials such as the products of the oxide-reduction and electro-refining processes. Copyright © 2016 Elsevier Ltd. All rights reserved.

Effect of reactor radiation on the thermal conductivity of TREAT fuel

NASA Astrophysics Data System (ADS)

Mo, Kun; Miao, Yinbin; Kontogeorgakos, Dimitrios C.; Connaway, Heather M.; Wright, Arthur E.; Yacout, Abdellatif M.

2017-04-01

The Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory is resuming operations after more than 20 years in latency in order to produce high-neutron-flux transients for investigating transient-induced behavior of reactor fuels and their interactions with other materials and structures. A parallel program is ongoing to develop a replacement core in which the fuel, historically containing highly-enriched uranium (HEU), is replaced by low-enriched uranium (LEU). Both the HEU and prospective LEU fuels are in the form of UO2 particles dispersed in a graphite matrix, but the LEU fuel will contain a much higher volume of UO2 particles, which may create a larger area of interphase boundaries between the particles and the graphite. This may lead to a higher volume fraction of graphite exposed to the fission fragments escaping from the UO2 particles, and thus may induce a higher volume of fission-fragment damage on the fuel graphite. In this work, we analyzed the reactor-radiation induced thermal conductivity degradation of graphite-based dispersion fuel. A semi-empirical method to model the relative thermal conductivity with reactor radiation was proposed and validated based on the available experimental data. Prediction of thermal conductivity degradation of LEU TREAT fuel during a long-term operation was performed, with a focus on the effect of UO2 particle size on fission-fragment damage. The proposed method can be further adjusted to evaluate the degradation of other properties of graphite-based dispersion fuel.

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

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

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

12. VIEW OF DEPLETED URANIUM INGOT AND MOLDS. DEPLETED URANIUM ...

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

12. VIEW OF DEPLETED URANIUM INGOT AND MOLDS. DEPLETED URANIUM CASTING OPERATIONS CEASED IN 1988. (11/14/57) - Rocky Flats Plant, Non-Nuclear Production Facility, South of Cottonwood Avenue, west of Seventh Avenue & east of Building 460, Golden, Jefferson County, CO

Critical experiments at Sandia National Laboratories : technical meeting on low-power critical facilities and small reactors.

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

Harms, Gary A.; Ford, John T.; Barber, Allison Delo

2010-11-01

Sandia National Laboratories (SNL) has conducted radiation effects testing for the Department of Energy (DOE) and other contractors supporting the DOE since the 1960's. Over this period, the research reactor facilities at Sandia have had a primary mission to provide appropriate nuclear radiation environments for radiation testing and qualification of electronic components and other devices. The current generation of reactors includes the Annular Core Research Reactor (ACRR), a water-moderated pool-type reactor, fueled by elements constructed from UO2-BeO ceramic fuel pellets, and the Sandia Pulse Reactor III (SPR-III), a bare metal fast burst reactor utilizing a uranium-molybdenum alloy fuel. The SPR-IIImore » is currently defueled. The SPR Facility (SPRF) has hosted a series of critical experiments. A purpose-built critical experiment was first operated at the SPRF in the late 1980's. This experiment, called the Space Nuclear Thermal Propulsion Critical Experiment (CX), was designed to explore the reactor physics of a nuclear thermal rocket motor. This experiment was fueled with highly-enriched uranium carbide fuel in annular water-moderated fuel elements. The experiment program was completed and the fuel for the experiment was moved off-site. A second critical experiment, the Burnup Credit Critical Experiment (BUCCX) was operated at Sandia in 2002. The critical assembly for this experiment was based on the assembly used in the CX modified to accommodate low-enriched pin-type fuel in water moderator. This experiment was designed as a platform in which the reactivity effects of specific fission product poisons could be measured. Experiments were carried out on rhodium, an important fission product poison. The fuel and assembly hardware for the BUCCX remains at Sandia and is available for future experimentation. The critical experiment currently in operation at the SPRF is the Seven Percent Critical Experiment (7uPCX). This experiment is designed to provide benchmark reactor physics data to support validation of the reactor physics codes used to design commercial reactor fuel elements in an enrichment range above the current 5% enrichment cap. A first set of critical experiments in the 7uPCX has been completed. More experiments are planned in the 7uPCX series. The critical experiments at Sandia National Laboratories are currently funded by the US Department of Energy Nuclear Criticality Safety Program (NCSP). The NCSP has committed to maintain the critical experiment capability at Sandia and to support the development of a critical experiments training course at the facility. The training course is intended to provide hands-on experiment experience for the training of new and re-training of practicing Nuclear Criticality Safety Engineers. The current plans are for the development of the course to continue through the first part of fiscal year 2011 with the development culminating is the delivery of a prototype of the course in the latter part of the fiscal year. The course will be available in fiscal year 2012.« less

Control console replacement at the WPI Reactor

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

Not Available

1992-01-01

With partial funding from the Department of Energy (DOE) University Reactor Instrumentation Upgrade Program (DOE Grant No. DE-FG02-90ER12982), the original control console at the Worcester Polytechnic Institute (WPI) Reactor has been replaced with a modern system. The new console maintains the original design bases and functionality while utilizing current technology. An advanced remote monitoring system has been added to augment the educational capabilities of the reactor. Designed and built by General Electric in 1959, the open pool nuclear training reactor at WPI was one of the first such facilities in the nation located on a university campus. Devoted to undergraduatemore » use, the reactor and its related facilities have been since used to train two generations of nuclear engineers and scientists for the nuclear industry. The reactor power level was upgraded from 1 to 10 kill in 1969, and its operating license was renewed for 20 years in 1983. In 1988, the reactor was converted to low enriched uranium. The low power output of the reactor and ergonomic facility design make it an ideal tool for undergraduate nuclear engineering education and other training.« less

The WPI reactor-readying for the next generation

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

Bobek, L.M.

1993-01-01

Built in 1959, the 10-kW open-pool nuclear training reactor at Worcester Polytechnic Institute (WPI) was one of the first such facilities in the nation located on a university campus. Since then, the reactor and its related facilities have been used to train two generations of nuclear engineers and scientists for the nuclear industry. With the use of nuclear technology playing an increasing role in many segments of the economy, WPI with its nuclear reactor facility is committed to continuing its mission of training future nuclear engineers and scientists. The WPI reactor includes a 6-in. beam port, graphite thermal column, andmore » in-core sample facility. The reactor, housed in an open 8000-gal tank of water, is designed so that the core is readily accessible. Both the control console and the peripheral counting equipment used for student projects and laboratory exercises are located in the reactor room. This arrangement provides convenience and flexibility in using the reactor for foil activations in neutron flux measurements, diffusion measurements, radioactive decay measurements, and the neutron activation of samples for analysis. In 1988, the reactor was successfully converted to low-enriched uranium fuel.« 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

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.

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.

Opportunities for Materials Science and Biological Research at the OPAL Research Reactor

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

Kennedy, S. J.

Neutron scattering techniques have evolved over more than 1/2 century into a powerful set of tools for determination of atomic and molecular structures. Modern facilities offer the possibility to determine complex structures over length scales from {approx}0.1 nm to {approx}500 nm. They can also provide information on atomic and molecular dynamics, on magnetic interactions and on the location and behaviour of hydrogen in a variety of materials. The OPAL Research Reactor is a 20 megawatt pool type reactor using low enriched uranium fuel, and cooled by water. OPAL is a multipurpose neutron factory with modern facilities for neutron beam research,more » radioisotope production and irradiation services. The neutron beam facility has been designed to compete with the best beam facilities in the world. After six years in construction, the reactor and neutron beam facilities are now being commissioned, and we will commence scientific experiments later this year. The presentation will include an outline of the strengths of neutron scattering and a description of the OPAL research reactor, with particular emphasis on it's scientific infrastructure. It will also provide an overview of the opportunities for research in materials science and biology that will be possible at OPAL, and mechanisms for accessing the facilities. The discussion will emphasize how researchers from around the world can utilize these exciting new facilities.« 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.

2004 Annual Health Physics Report for the HEU Transparency Program

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

Radev, R

2005-04-01

During the 2004 calendar year, LLNL provided health physics support for the Highly Enriched Uranium Transparency Implementation Program (HEU-TIP) in external and internal radiation protection and technical expertise into matters related to BDMS radioactive sources and Russian radiation safety regulatory compliance. For the calendar year 2004, there were 200 person-trips that required dose monitoring of the U.S. monitors. Of the 200 person-trips, 183 person-trips were SMVs and 17 person-trips were Transparency Monitoring Office (TMO) trips. Eight person-trips from the SMV trips were continuation trips of TMO monitors to facilities other than UEIP. The LLNL Safety Laboratories' Division provided the dosimetrymore » services for the HEU-TIP monitors.« less

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...

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...

77 FR 70486 - Supplemental Environmental Impact Statement for Proposed Dewey-Burdock In-Situ Uranium Recovery...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-11-26

... Proposed Dewey- Burdock In-Situ Uranium Recovery Project in Custer and Fall River Counties, SD AGENCY... draft Supplemental Environmental Impact Statement (Draft SEIS) for the Dewey-Burdock In-Situ Uranium... NUREG-1910, ``Generic Environmental Impact Statement for In-Situ Leach Uranium Milling Facilities,'' May...

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

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

''Talk To Me!'' - Lessons Learned In Communicating Risks To Tenants And Others Involved In The Department Of Energy's Reindustrialization Program

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

Cusick, Lesley T.; Golden, Karen M.

2003-02-26

Communicating risk information is more difficult than assessing it. The latter relies on data, formulas, theorems and mathematical relationships that, with some effort, can be logically explained to another person; it's objective. Communicating risks, however, is subjective and relies on personalities, perceptions and predisposition, as well as emotions. Most notably the emotion is fear--fear of the unknown, fear of the message, the messenger, or the impact of the information on something of value to the person asking the questions. The Department of Energy's Oak Ridge Operations Office is engaged in a Reindustrialization program to lease (and most recently, to transfer)more » formerly used facilities to private sector entities. The facilities are located at the East Tennessee Technology Park, originally a gaseous diffusion plant operated to enrich uranium for World War II efforts and later for use as fuel in civilian nuclear reactors.« less

Special Analysis: 2017-001 Disposal of Drums Containing Enriched Uranium in Pit 38 at Technical Area 54, Area G

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

Birdsell, Kay Hanson; Stauffer, Philip H.; French, Sean B.

Los Alamos National Laboratory (LANL) generates radioactive waste as a result of various activities. Operational waste is generated from a wide variety of research and development activities including nuclear weapons development, energy production, and medical research. Environmental restoration (ER), and decontamination and decommissioning (D&D) waste is generated as contaminated sites and facilities at LANL undergo cleanup or remediation. The majority of this waste is low-level radioactive waste (LLW) and is disposed of at the Technical Area 54 (TA-54), Area G disposal facility. This special analysis, SA 2017-001, evaluates the potential impacts of disposing of this waste in Pit 38 atmore » Area G based on the assumptions that form the basis of the Area G PA/CA. Section 2 describes the methods used to conduct the analysis; the results of the evaluation are provided in Section 3; and conclusions and recommendations are provided in Section 4.« less

Project C-018H, 242-A Evaporator/PUREX Plant Process Condensate Treatment Facility, functional design criteria. Revision 3

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

Sullivan, N.

1995-05-02

This document provides the Functional Design Criteria (FDC) for Project C-018H, the 242-A Evaporator and Plutonium-Uranium Extraction (PUREX) Plant Condensate Treatment Facility (Also referred to as the 200 Area Effluent Treatment Facility [ETF]). The project will provide the facilities to treat and dispose of the 242-A Evaporator process condensate (PC), the Plutonium-Uranium Extraction (PUREX) Plant process condensate (PDD), and the PUREX Plant ammonia scrubber distillate (ASD).

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

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

Parker, Frank L.

2012-07-01

Current fuel cycles offer a number of opportunities for access to plutonium, opportunities to create highly enriched uranium and access highly radioactive wastes to create nuclear weapons and 'dirty' bombs. The non-proliferating fuel cycle however eliminates or reduces such opportunities and access by eliminating the mining, milling and enrichment of uranium. The non-proliferating fuel cycle also reduces the production of plutonium per unit of energy created, eliminates reprocessing and the separation of plutonium from the spent fuel and the creation of a stream of high-level waste. It further simplifies the search for land based deep geologic repositories and interim storagemore » sites for spent fuel in the USA by disposing of the spent fuel in deep sub-seabed sediments after storing the spent fuel at U.S. Navy Nuclear Shipyards that have the space and all of the necessary equipment and security already in place. The non-proliferating fuel cycle also reduces transportation risks by utilizing barges for the collection of spent fuel and transport to the Navy shipyards and specially designed ships to take the spent fuel to designated disposal sites at sea and to dispose of them there in deep sub-seabed sediments. Disposal in the sub-seabed sediments practically eliminates human intrusion. Potential disposal sites include Great Meteor East and Southern Nares Abyssal Plain. Such sites then could easily become international disposal sites since they occur in the open ocean. It also reduces the level of human exposure in case of failure because of the large physical and chemical dilution and the elimination of a major pathway to man-seawater is not potable. Of course, the recovery of uranium from sea water and the disposal of spent fuel in sub-seabed sediments must be proven on an industrial scale. All other technologies are already operating on an industrial scale. If externalities, such as reduced terrorist threats, environmental damage (including embedded emissions), long term care, reduced access to 'dirty' bomb materials, the social and political costs of siting new facilities and the psychological impact of no solution to the nuclear waste problem, were taken into account, the costs would be far lower than those of the present fuel cycle. (authors)« 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).

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

Final report of the decontamination and decommissioning of the exterior land areas at the Grand Junction Projects Office facility

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

Widdop, M.R.

1995-09-01

The US Department of Energy (DOE) Grand Junction Projects Office (GJPO) facility occupies approximately 56.4 acres (22.8 hectares) along the Gunnison River near Grand Junction, Colorado. The site was contaminated with uranium ore and mill tailings during uranium-refining activities conducted by the Manhattan Engineer District and during pilot-milling experiments conducted for the US Atomic Energy Commission`s (AEC`s) domestic uranium procurement program. The GJPO facility was the collection and assay point for AEC uranium and vanadium oxide purchases until the early 1970s. The DOE Decontamination and Decommissioning Program sponsored the Grand Junction Projects Office Remedial Action Project (GJPORAP) to remediate themore » facility lands, site improvements, and the underlying aquifer. The site contractor, Rust Geotech, was the Remedial Action Contractor for GJPORAP. The exterior land areas of the facility assessed as contaminated have been remediated in accordance with identified standards and can be released for unrestricted use. Restoration of the aquifer will be accomplished through the natural flushing action of the aquifer during the next 50 to 80 years. The remediation of the DOE-GJPO facility buildings is ongoing and will be described in a separate report.« 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.

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.

PDRD (SR13046) TRITIUM PRODUCTION FINAL REPORT

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

Smith, P.; Sheetz, S.

Utilizing the results of Texas A&M University (TAMU) senior design projects on tritium production in four different small modular reactors (SMR), the Savannah River National Laboratory’s (SRNL) developed an optimization model evaluating tritium production versus uranium utilization under a FY2013 plant directed research development (PDRD) project. The model is a tool that can evaluate varying scenarios and various reactor designs to maximize the production of tritium per unit of unobligated United States (US) origin uranium that is in limited supply. The primary module in the model compares the consumption of uranium for various production reactors against the base case ofmore » Watts Bar I running a nominal load of 1,696 tritium producing burnable absorber rods (TPBARs) with an average refueling of 41,000 kg low enriched uranium (LEU) on an 18 month cycle. After inputting an initial year, starting inventory of unobligated uranium and tritium production forecast, the model will compare and contrast the depletion rate of the LEU between the entered alternatives. This is an annual tritium production rate of approximately 0.059 grams of tritium per kilogram of LEU (g-T/kg-LEU). To date, the Nuclear Regulatory Commission (NRC) license has not been amended to accept a full load of TPBARs so the nominal tritium production has not yet been achieved. The alternatives currently loaded into the model include the three light water SMRs evaluated in TAMU senior projects including, mPower, Holtec and NuScale designs. Initial evaluations of tritium production in light water reactor (LWR) based SMRs using optimized loads TPBARs is on the order 0.02-0.06 grams of tritium per kilogram of LEU used. The TAMU students also chose to model tritium production in the GE-Hitachi SPRISM, a pooltype sodium fast reactor (SFR) utilizing a modified TPBAR type target. The team was unable to complete their project so no data is available. In order to include results from a fast reactor, the SRNL Technical Advisory Committee (TAC) ran a Monte Carlo N-Particle (MCNP) model of a basic SFR for comparison. A 600MWth core surrounded by a lithium blanket produced approximately 1,000 grams of tritium annually with a 13% enriched, 6 year core. This is similar results to a mid-1990’s study where the Fast Flux Test Facility (FFTF), a 400 MWth reactor at the Idaho National Laboratory (INL), could produce about 1,000 grams with an external lithium target. Normalized to the LWRs values, comparative tritium production for an SFR could be approximately 0.31 g-T/kg LEU.« less

Effect of reactor radiation on the thermal conductivity of TREAT fuel

DOE PAGES

Mo, Kun; Miao, Yinbin; Kontogeorgakos, Dimitrios C.; ...

2017-02-04

The Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory is resuming operations after more than 20 years in latency in order to produce high-neutron-flux transients for investigating transient-induced behavior of reactor fuels and their interactions with other materials and structures. A parallel program is ongoing to develop a replacement core in which the fuel, historically containing highly-enriched uranium (HEU), is replaced by low-enriched uranium (LEU). Both the HEU and prospective LEU fuels are in the form of UO 2 particles dispersed in a graphite matrix, but the LEU fuel will contain a much higher volume of UO 2more » particles, which may create a larger area of interphase boundaries between the particles and the graphite. This may lead to a higher volume fraction of graphite exposed to the fission fragments escaping from the UO 2 particles, and thus may induce a higher volume of fission-fragment damage on the fuel graphite. In this work, we analyzed the reactor-radiation induced thermal conductivity degradation of graphite-based dispersion fuel. A semi-empirical method to model the relative thermal conductivity with reactor radiation was proposed and validated based on the available experimental data. Prediction of thermal conductivity degradation of LEU TREAT fuel during a long-term operation was performed, with a focus on the effect of UO 2 particle size on fission-fragment damage. Lastly, the proposed method can be further adjusted to evaluate the degradation of other properties of graphite-based dispersion fuel.« less

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

Margaret A. Marshall; John D. Bess; Yevgeniy Rozhikhin

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 attempt to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s (HEU-MET-FAST-001). 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 withmore » the GODIVA I experiments[1]. Part of the experimental series was the measurement of the delayed neutron fraction, ßeff, using time correlation measurements and using the central void reactivity measurement. The time correlations measurements were rejected by the experimenter. The measurements using the central void reactivity measurement yielded a ßeff value of 0.00657, which agrees well with the value measured with GODIVA I (0.0066). This measurement is evaluated, found to be acceptable, and discussed in extensive detail in “ORSphere: Physics Measurements for Bare, HEU(93.2) Metal Sphere”[2]. In order to determine the delayed neutron fraction using the central void reactivity delayed neutron parameters must be used. The experimenter utilized the delayed neutron parameters set forth by Keepin, Wimment, and Zeigler[3]. If the derivation of the ßeff is repeated with different delayed neutron parameters from various modern nuclear data sets the resulting values vary greatly from the expected results.« less

Effect of reactor radiation on the thermal conductivity of TREAT fuel

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

Mo, Kun; Miao, Yinbin; Kontogeorgakos, Dimitrios C.

The Transient Reactor Test Facility (TREAT) at the Idaho National Laboratory is resuming operations after more than 20 years in latency in order to produce high-neutron-flux transients for investigating transient-induced behavior of reactor fuels and their interactions with other materials and structures. A parallel program is ongoing to develop a replacement core in which the fuel, historically containing highly-enriched uranium (HEU), is replaced by low-enriched uranium (LEU). Both the HEU and prospective LEU fuels are in the form of UO 2 particles dispersed in a graphite matrix, but the LEU fuel will contain a much higher volume of UO 2more » particles, which may create a larger area of interphase boundaries between the particles and the graphite. This may lead to a higher volume fraction of graphite exposed to the fission fragments escaping from the UO 2 particles, and thus may induce a higher volume of fission-fragment damage on the fuel graphite. In this work, we analyzed the reactor-radiation induced thermal conductivity degradation of graphite-based dispersion fuel. A semi-empirical method to model the relative thermal conductivity with reactor radiation was proposed and validated based on the available experimental data. Prediction of thermal conductivity degradation of LEU TREAT fuel during a long-term operation was performed, with a focus on the effect of UO 2 particle size on fission-fragment damage. Lastly, the proposed method can be further adjusted to evaluate the degradation of other properties of graphite-based dispersion fuel.« 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...

76 FR 37160 - Notice of Availability of Final Supplemental Environmental Impact Statement for the Lost Creek In...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-06-24

... Environmental Impact Statement for the Lost Creek In-Situ Recovery (ISR) Project in Sweetwater County, WY; Supplement to the Generic Environmental Impact Statement for In-Situ Leach Uranium Milling Facilities AGENCY... Statement (SEIS) to the Generic Environmental Impact Statement for In- Situ Leach Uranium Milling Facilities...

75 FR 13141 - Powertech (USA), Inc.; Establishment of Atomic Safety and Licensing Board

Federal Register 2010, 2011, 2012, 2013, 2014

2010-03-18

... following proceeding: Powertech (USA) Inc. (Dewey-Burdock In Situ Uranium Recovery Facility). This Board is... Powertech (USA) Inc.'s application for a source materials license for an in situ uranium recovery facility... Commission dated December 29, 1972, published in the Federal Register, 37 FR 28,710 (1972), and the...

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.

Molybdenum-UO2 cermet irradiation at 1145 K.

NASA Technical Reports Server (NTRS)

Mcdonald, G.

1971-01-01

Two molybdenum-uranium dioxide cermet fuel pins with molybdenum clad were fission-heated in a forced-convection helium coolant for sufficient time to achieve 5.3% burnup. The cermet core contained 20 wt % of 93.2% enriched uranium dioxide. The results were as follows: there was no visible change in the appearance of the molybdenum clad during irradiation; the maximum increase in diameter of the fuel pins was 0.8%; there was no migration of uranium dioxide along grain boundaries and no evident interaction between molybdenum and uranium dioxide; and, finally, approximately 12% of the fission gas formed was released from the cermet core into the gas plenum.

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...

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

Integrating UF6 Cylinder RF Tracking With Continuous Load Cell Monitoring for Verifying Declared UF6 Feed and Withdrawal Operations Verifying Declared UF6 Feed and Withdrawal Operations

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

Krichinsky, Alan M; Miller, Paul; Pickett, Chris A

2009-01-01

Oak Ridge National Laboratory is demonstrating the integration of UF6 cylinder tracking, using RF technology, with continuous load cell monitoring (CLCM) at mock UF6 feed and withdrawal (F&W) stations. CLCM and cylinder tracking are two of several continuous-monitoring technologies that show promise in providing integrated safeguards of F&W operations at enrichment plants. Integrating different monitoring technologies allows advanced, automated event processing to screen innocuous events thereby minimizing false alerts to independent inspectors. Traditionally, international inspectors rely on batch verification of material inputs and outputs derived from operator declarations and periodic on-site inspections at uranium enrichment plants or other nuclear processingmore » facilities. Continuously monitoring F&W activities between inspections while providing filtered alerts of significant operational events will substantially increase the amount of valuable information available to inspectors thereby promising to enhance the effectiveness of safeguards and to improve efficiency in conducting on-site inspections especially at large plants for ensuring that all operations are declared.« less

75 FR 62153 - Notice of the Nuclear Regulatory Commission Issuance of Materials License SUA-1596 for Uranium...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-10-07

... Commission Issuance of Materials License SUA-1596 for Uranium One Americas, Inc. Moore Ranch In Situ Recovery..., Inc. (Uranium One) for its Moore Ranch uranium in situ recovery (ISR) facility in Campbell County... discussed in detail were the applicant's proposal as described in its license application to conduct in situ...

Neutronics qualification of the Jules Horowitz reactor fuel by interpretation of the VALMONT experimental program - Transposition of the uncertainties on the reactivity of JHR with JEF2.2 and JEFF3.1.1

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

Leray, O.; Hudelot, J. P.; Antony, M.

2011-07-01

The new European material testing Jules Horowitz Reactor (JHR), currently under construction in Cadarache center (CEA France), will use LEU (20% enrichment in {sup 235}U) fuels (U{sub 3}Si{sub 2} for the start up and UMoAl in the future) which are quite different from the industrial oxide fuel, for which an extensive neutronics qualification database has been established. The HORUS3D/N neutronics calculation scheme, used for the design and safety studies of the JHR, is being developed within the framework of a rigorous verification-validation-qualification methodology. In this framework, the experimental VALMONT (Validation of Aluminium Molybdenum uranium fuel for Neutronics) program has beenmore » performed in the MINERVE facility of CEA Cadarache (France), in order to qualify the capability of HORUS3D/N to accurately calculate the reactivity of the JHR reactor. The MINERVE facility using the oscillation technique provides accurate measurements of reactivity effect of samples. The VALMONT program includes oscillations of samples of UAl{sub x}/Al and UMo/Al with enrichments ranging from 0.2% to 20% and Uranium densities from 2.2 to 8 g/cm{sup 3}. The geometry of the samples and the pitch of the experimental lattice ensure maximum representativeness with the neutron spectrum expected for JHR. By comparing the effect of the sample with the one of a known fuel specimen, the reactivity effect can be measured in absolute terms and be compared to computational results. Special attention was paid to the rigorous determination and reduction of the experimental uncertainties. The calculational analysis of the VALMONT results was performed with the French deterministic code APOLLO2. A comparison of the impact of the different calculation methods, data libraries and energy meshes that were tested is presented. The interpretation of the VALMONT experimental program allowed the qualification of JHR fuel UMoAl8 (with an enrichment of 19.75% {sup 235}U) by the Minerve-dedicated interpretation tool: PIMS. The effect of energy meshes and evaluations put forward the JEFF3.1.1/SHEM scheme that leads to a better calculation of the reactivity effect of VALMONT samples. Then, in order to quantify the impact of the uncertainties linked to the basic nuclear data, their propagation from the cross section measurement to the final computational result was analysed in a rigorous way by using a nuclear data re-estimation method based on Gauss-Newton iterations. This study concludes that the prior uncertainties due to nuclear data (uranium, aluminium, beryllium and water) on the reactivity of the Begin Of Cycle (BOC) for the JHR core reach 1217 pcm at 2{sigma}. Now, the uppermost uncertainty on the JHR reactivity is due to aluminium. (authors)« less

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.

ZPPR-20 phase D : a cylindrical assembly of polyethylene moderated U metal reflected by beryllium oxide and polyethylene.

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

Lell, R.; Grimm, K.; McKnight, R.

The Zero Power Physics Reactor (ZPPR) fast critical facility was built at the Argonne National Laboratory-West (ANL-W) site in Idaho in 1969 to obtain neutron physics information necessary for the design of fast breeder reactors. The ZPPR-20D Benchmark Assembly was part of a series of cores built in Assembly 20 (References 1 through 3) of the ZPPR facility to provide data for developing a nuclear power source for space applications (SP-100). The assemblies were beryllium oxide reflected and had core fuel compositions containing enriched uranium fuel, niobium and rhenium. ZPPR-20 Phase C (HEU-MET-FAST-075) was built as the reference flight configuration.more » Two other configurations, Phases D and E, simulated accident scenarios. Phase D modeled the water immersion scenario during a launch accident, and Phase E (SUB-HEU-MET-FAST-001) modeled the earth burial scenario during a launch accident. Two configurations were recorded for the simulated water immersion accident scenario (Phase D); the critical configuration, documented here, and the subcritical configuration (SUB-HEU-MET-MIXED-001). Experiments in Assembly 20 Phases 20A through 20F were performed in 1988. The reference water immersion configuration for the ZPPR-20D assembly was obtained as reactor loading 129 on October 7, 1988 with a fissile mass of 167.477 kg and a reactivity of -4.626 {+-} 0.044{cents} (k {approx} 0.9997). The SP-100 core was to be constructed of highly enriched uranium nitride, niobium, rhenium and depleted lithium. The core design called for two enrichment zones with niobium-1% zirconium alloy fuel cladding and core structure. Rhenium was to be used as a fuel pin liner to provide shut down in the event of water immersion and flooding. The core coolant was to be depleted lithium metal ({sup 7}Li). The core was to be surrounded radially with a niobium reactor vessel and bypass which would carry the lithium coolant to the forward inlet plenum. Immediately inside the reactor vessel was a rhenium baffle which would act as a neutron curtain in the event of water immersion. A fission gas plenum and coolant inlet plenum were located axially forward of the core. Some material substitutions had to be made in mocking up the SP-100 design. The ZPPR-20 critical assemblies were fueled by 93% enriched uranium metal because uranium nitride, which was the SP-100 fuel type, was not available. ZPPR Assembly 20D was designed to simulate a water immersion accident. The water was simulated by polyethylene (CH{sub 2}), which contains a similar amount of hydrogen and has a similar density. A very accurate transformation to a simplified model is needed to make any of the ZPPR assemblies a practical criticality-safety benchmark. There is simply too much geometric detail in an exact model of a ZPPR assembly, particularly as complicated an assembly as ZPPR-20D. The transformation must reduce the detail to a practical level without masking any of the important features of the critical experiment. And it must do this without increasing the total uncertainty far beyond that of the original experiment. Such a transformation will be described in a later section. First, Assembly 20D was modeled in full detail--every plate, drawer, matrix tube, and air gap was modeled explicitly. Then the regionwise compositions and volumes from this model were converted to an RZ model. ZPPR Assembly 20D has been determined to be an acceptable criticality-safety benchmark experiment.« less

Safety approaches for high power modular laser operation

NASA Astrophysics Data System (ADS)

Handren, R. T.

1993-03-01

Approximately 20 years ago, a program was initiated at the Lawrence Livermore National Laboratory (LLNL) to study the feasibility of using lasers to separate isotopes of uranium and other materials. Of particular interest was the development of a uranium enrichment method for the production of commercial nuclear power reactor fuel to replace current more expensive methods. The Uranium Atomic Vapor Laser Isotope Separation (U-AVLIS) Program progressed to the point where a plant-scale facility to demonstrate commercial feasibility was built and is being tested. The U-AVLIS Program uses copper vapor lasers which pump frequency selective dye lasers to photoionize uranium vapor produced by an electron beam. The selectively ionized isotopes are electrostatically collected. The copper lasers are arranged in oscillator/amplifier chains. The current configuration consists of 12 chains, each with a nominal output of 800 W for a system output in excess of 9 kW. The system requirements are for continuous operation (24 h a day, 7 days a week) and high availability. To meet these requirements, the lasers are designed in a modular form allowing for rapid change-out of the lasers requiring maintenance. Since beginning operation in early 1985, the copper lasers have accumulated over 2 million unit hours at a greater than 90% availability. The dye laser system provides approximately 2.5 kW average power in the visible wavelength range. This large-scale laser system has many safety considerations, including high-power laser beams, high voltage, and large quantities (approximately 3000 gal) of ethanol dye solutions. The Laboratory's safety policy requires that safety controls be designed into any process, equipment, or apparatus in the form of engineering controls. Administrative controls further reduce the risk to an acceptable level. Selected examples of engineering and administrative controls currently being used in the U-AVLIS Program are described.

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

Parameter Study of the LIFE Engine Nuclear Design

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

Kramer, K J; Meier, W R; Latkowski, J F

2009-07-10

LLNL is developing the nuclear fusion based Laser Inertial Fusion Energy (LIFE) power plant concept. The baseline design uses a depleted uranium (DU) fission fuel blanket with a flowing molten salt coolant (flibe) that also breeds the tritium needed to sustain the fusion energy source. Indirect drive targets, similar to those that will be demonstrated on the National Ignition Facility (NIF), are ignited at {approx}13 Hz providing a 500 MW fusion source. The DU is in the form of a uranium oxycarbide kernel in modified TRISO-like fuel particles distributed in a carbon matrix forming 2-cm-diameter pebbles. The thermal power ismore » held at 2000 MW by continuously varying the 6Li enrichment in the coolants. There are many options to be considered in the engine design including target yield, U-to-C ratio in the fuel, fission blanket thickness, etc. Here we report results of design variations and compare them in terms of various figures of merit such as time to reach a desired burnup, full-power years of operation, time and maximum burnup at power ramp down and the overall balance of plant utilization.« less

Counting neutrons from the spontaneous fission of {sup 238}U using scintillation detectors and mixed field analysers

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

Parker, Helen M. O'D.; Joyce, Malcolm J.; Jones, Ashley

2015-07-01

It is well documented that {sup 238}U decays by spontaneous fission, and that it is the main component of most nuclear fuels. As nuclear fuels are largely classed as Special Nuclear Material (SNM), they have to be fully accounted for by owners and processing facilities. One possible method for verifying declared amounts of SNM is to count the spontaneous neutrons produced from {sup 238}U. Using four EJ-309 liquid scintillation detectors and a mixed field analyser, spontaneous neutrons from 16.4 g of depleted uranium (0.3% enrichment) have been assayed. The assay method shows promising results and this proof of principle willmore » be researched further in order for it to be applied in an industrial setting. (authors)« less

2007 Annual Health Physics Report for the HEU Transparency Program

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

Radev, R

2008-04-09

During the 2007 calendar year, Lawrence Livermore National Laboratory (LLNL) provided health physics support for the Highly Enriched Uranium (HEU) Transparency Program for external and internal radiation protection and technical expertise related to BDMS radioactive sources and Russian radiation safety regulatory compliance. For the calendar year 2007, there were 172 person-trips that required dose monitoring of the U.S. monitors. Of the 172 person-trips, 160 person-trips were SMVs and 12 person-trips were Transparency Monitoring Office (TMO) trips. There were 12 monitoring visits by TMO monitors to facilities other than UEIE and 10 to UEIE itself. There were two monitoring visits (sourcemore » changes) that were back to back with 14 monitors. LLNL's Hazard Control Division laboratories provided the dosimetry services for the HEU Transparency monitors.« less

FastDart : a fast, accurate and friendly version of DART code.

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

Rest, J.; Taboada, H.

2000-11-08

A new enhanced, visual version of DART code is presented. DART is a mechanistic model based code, developed for the performance calculation and assessment of aluminum dispersion fuel. Major issues of this new version are the development of a new, time saving calculation routine, able to be run on PC, a friendly visual input interface and a plotting facility. This version, available for silicide and U-Mo fuels,adds to the classical accuracy of DART models for fuel performance prediction, a faster execution and visual interfaces. It is part of a collaboration agreement between ANL and CNEA in the area of Lowmore » Enriched Uranium Advanced Fuels, held by the Implementation Arrangement for Technical Exchange and Cooperation in the Area of Peaceful Uses of Nuclear Energy.« less

Uranium production in Romania

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

NONE

1994-12-01

This article reviews uranium production in Romania. Geological aspects of the country are discussed, and known uranium deposits are noted. Uranium mining and milling activities are also covered. Utilization of Romania`s uranium production industry will primarily be to supply the country`s nuclear power program, and with the present adequate supplies and the operation of their recently revamped fuel production facility, Romania should be self-reliant in the front end of the nuclear fuel cycle.

Control console replacement at the WPI Reactor. [Final report

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

Not Available

1992-12-31

With partial funding from the Department of Energy (DOE) University Reactor Instrumentation Upgrade Program (DOE Grant No. DE-FG02-90ER12982), the original control console at the Worcester Polytechnic Institute (WPI) Reactor has been replaced with a modern system. The new console maintains the original design bases and functionality while utilizing current technology. An advanced remote monitoring system has been added to augment the educational capabilities of the reactor. Designed and built by General Electric in 1959, the open pool nuclear training reactor at WPI was one of the first such facilities in the nation located on a university campus. Devoted to undergraduatemore » use, the reactor and its related facilities have been since used to train two generations of nuclear engineers and scientists for the nuclear industry. The reactor power level was upgraded from 1 to 10 kill in 1969, and its operating license was renewed for 20 years in 1983. In 1988, the reactor was converted to low enriched uranium. The low power output of the reactor and ergonomic facility design make it an ideal tool for undergraduate nuclear engineering education and other training.« 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.

75 FR 467 - Notice of Opportunity for Hearing, License Application Request of Powertech (USA) Inc. Dewey...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-01-05

... Hearing, License Application Request of Powertech (USA) Inc. Dewey-Burdock In Situ Uranium Recovery... Regulatory Commission (NRC) for the Dewey-Burdock In Situ Recovery Facility in Fall River and Custer Counties, South Dakota. The Dewey-Burdock facility would involve the recovery of uranium by in situ recovery (ISR...

Depleted UF6 Internet Resources

Science.gov Websites

been used to color glass for almost 2 millennia. A uranium-colored glass object was found near Naples , Italy, and dated to about 79 A.D. Uranium oxide added to glass produces a yellow to greenish hue. more Board Defense Nuclear Facilities Safety Board (DNFSB) The Defense Nuclear Facilities Safety Board

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.

Laboratory instrumentation modernization at the WPI Nuclear Reactor Facility

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

Not Available

1995-01-01

With partial funding from the Department of Energy (DOE) University Reactor Instrumentation Program several laboratory instruments utilized by students and researchers at the WPI Nuclear Reactor Facility have been upgraded or replaced. Designed and built by General Electric in 1959, the open pool nuclear training reactor at WPI was one of the first such facilities in the nation located on a university campus. Devoted to undergraduate use, the reactor and its related facilities have been since used to train two generations of nuclear engineers and scientists for the nuclear industry. The low power output of the reactor and an ergonomicmore » facility design make it an ideal tool for undergraduate nuclear engineering education and other training. The reactor, its control system, and the associate laboratory equipment are all located in the same room. Over the years, several important milestones have taken place at the WPI reactor. In 1969, the reactor power level was upgraded from 1 kW to 10 kW. The reactor`s Nuclear Regulatory Commission operating license was renewed for 20 years in 1983. In 1988, under DOE Grant No. DE-FG07-86ER75271, the reactor was converted to low-enriched uranium fuel. In 1992, again with partial funding from DOE (Grant No. DE-FG02-90ER12982), the original control console was replaced.« less

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

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

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

Neutron-rich isotope production using a uranium carbide - carbon nanotubes SPES target prototype

NASA Astrophysics Data System (ADS)

Corradetti, S.; Biasetto, L.; Manzolaro, M.; Scarpa, D.; Carturan, S.; Andrighetto, A.; Prete, G.; Vasquez, J.; Zanonato, P.; Colombo, P.; Jost, C. U.; Stracener, D. W.

2013-05-01

The SPES (Selective Production of Exotic Species) project, under development at the Istituto Nazionale di Fisica Nucleare - Laboratori Nazionali di Legnaro (INFN-LNL), is a new-generation Isotope Separation On-Line (ISOL) facility for the production of radioactive ion beams by means of the proton-induced fission of uranium. In the framework of the research on the SPES target, seven uranium carbide discs, obtained by reacting uranium oxide with graphite and carbon nanotubes, were irradiated with protons at the Holifield Radioactive Ion Beam Facility (HRIBF) of Oak Ridge National Laboratory (ORNL). In the following, the yields of several fission products obtained during the experiment are presented and discussed. The experimental results are then compared to those obtained using a standard uranium carbide target. The reported data highlights the capability of the new type of SPES target to produce and release isotopes of interest for the nuclear physics community.

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

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.

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.

High frequency EMI sensing for estimating depleted uranium radiation levels in soil

NASA Astrophysics Data System (ADS)

Shubitidze, Fridon; Barrowes, Benjamin E.; Ballard, John; Unz, Ron; Randle, Adam; Larson, Steve L.; O'Neill, Kevin A.

2018-04-01

This paper studies high (100 kHz up to 15 MHz) frequency electromagnetic responses (HFEMI) for DU metallic pieces and DU contaminated soils and derives a simple empirical expression from the measured HFEMI data for estimating DU contamination levels in soil. Depleted uranium (DU) is the byproduct of uranium enrichment and contains 33% less radioactive isotopes than natural uranium. There are at least thirty facilities at fourteen separate locations in the US, where munitions containing DU have been evaluated or used for training. At these sites, which vary in size, evaluation studies have been conducted with and without catch boxes. In addition, the DoD used DU at open firing ranges as large as thousands of acres (hundreds of hectares), for both artillery and aircraft training. These activities have left a legacy of DU contamination. Currently at military sites where DU munitions have been or are being used, cleanup activities mainly are done by excavating and shipping large volumes of site soil and berm materials to a hazardous material radiation disposal site. This approach is very time consuming, costly, and associated with the potential for exposure of personnel performing excavation and transportation. It also limits range use during the operation. So, there is an urgent need for technologies for rapid surveying of large areas to detect, locate, and removal of DU contaminants at test sites. Additionally, the technologies are needed to detect material at a depth of at least 30 cm as well as discriminate between DU metals and oxides from natural uranium and from other conductive metals such as natural and man-made range clutter. One of the potential technologies for estimating DU radiation levels in soils is HFEMI sensing. In this paper, HFEMI signals are collected for DU metal pieces, sodium diunarate (Na2U2 O3) and tri-uranium octoxide (U3O8). The EMI signal's sensitivity with respect to DU material composition and conditions are illustrated and analyzed. A new scheme for extracting near-surface soil's EM parameters is formulated.

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.

China and Proliferation of Weapons of Mass Destruction and Missiles: Policy Issues

DTIC Science & Technology

2014-01-03

countries) for secret nuclear weapons facilities, while experts from China worked at a uranium mine at Saghand and a centrifuge facility (for uranium...declaration from North Korea for outside verification. 89 Barbara Opall -Rome and...that the China Guangfa Bank engaged in business with the DPRK’s arms dealer, Global Trading and Technology (a front for Korea Mining Development

75 FR 3261 - Powertech (USA) Inc.; Dewey-Burdock Project; New Source Material License Application; Notice of...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-01-20

... in-situ uranium recovery (ISR, also known as in-situ leach) facilities, and would require restoration... Environmental Impact Statement for In-Situ Leach Uranium Milling Facilities (ISR GEIS) that was published in... published in the Federal Register on January 05, 2010 (75 FR 467-471). The purpose of this notice of intent...

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.

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

Medical Isotope Production Analyses In KIPT Neutron Source Facility

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

Talamo, Alberto; Gohar, Yousry

Medical isotope production analyses in Kharkov Institute of Physics and Technology (KIPT) neutron source facility were performed to include the details of the irradiation cassette and the self-shielding effect. An updated detailed model of the facility was used for the analyses. The facility consists of an accelerator-driven system (ADS), which has a subcritical assembly using low-enriched uranium fuel elements with a beryllium-graphite reflector. The beryllium assemblies of the reflector have the same outer geometry as the fuel elements, which permits loading the subcritical assembly with different number of fuel elements without impacting the reflector performance. The subcritical assembly is drivenmore » by an external neutron source generated from the interaction of 100-kW electron beam with a tungsten target. The facility construction was completed at the end of 2015, and it is planned to start the operation during the year of 2016. It is the first ADS in the world, which has a coolant system for removing the generated fission power. Argonne National Laboratory has developed the design concept and performed extensive design analyses for the facility including its utilization for the production of different radioactive medical isotopes. 99Mo is the parent isotope of 99mTc, which is the most commonly used medical radioactive isotope. Detailed analyses were performed to define the optimal sample irradiation location and the generated activity, for several radioactive medical isotopes, as a function of the irradiation time.« less

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

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.