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1

Highly enriched uranium (HEU) storage and disposition program plan  

SciTech Connect

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

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

1995-01-01T23:59:59.000Z

2

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

SciTech Connect

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

Swindle, D.W.

1990-03-01T23:59:59.000Z

3

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

SciTech Connect

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

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

1987-01-01T23:59:59.000Z

4

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

SciTech Connect

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

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

2000-04-07T23:59:59.000Z

5

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

SciTech Connect

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

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

1998-08-01T23:59:59.000Z

6

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

SciTech Connect

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

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

2001-01-01T23:59:59.000Z

7

Active Interrogation Observables for Enrichment Determination of DU Shielded HEU Metal Assemblies with Limited Geometrical Information  

SciTech Connect

Determining the enrichment of highly enriched uranium (HEU) metal assemblies shielded by depleted uranium (DU) proves a unique challenge to currently employed measurement techniques. Efforts to match time-correlated neutron distributions obtained through active interrogation to Monte Carlo simulations of the assemblies have shown promising results, given that the exact geometries of both the HEU metal assemblies and DU shields are known from imaging and fission site mapping. In certain situations, however, it is desirable to obtain enrichment with limited or no geometrical information of the assemblies being measured. This paper explores the possibility that the utilization of observables in the interrogation of assemblies by time-tagged D-T neutrons, including time-correlated distribution of neutrons and gammas using liquid scintillators operating on the fission chain time scale, can lead to enrichment determination without a complete set of geometrical information.

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

2011-01-01T23:59:59.000Z

8

Unattended Environmental Sampling and Laser-based Enrichment Assay for Detection of Undeclared HEU Production in Enrichment Plants  

Science Conference Proceedings (OSTI)

Nuclear power is enjoying rapid growth as government energy policies and public demand shift toward carbon neutral energy production. Accompanying the growth in nuclear power is the requirement for increased nuclear fuel production, including a significant expansion in uranium enrichment capacity. Essential to the success of the nuclear energy renaissance is the development and implementation of sustainable, proliferation-resistant nuclear power generation. Unauthorized production of highly enriched uranium (HEU) remains the primary proliferation concern for modern gaseous centrifuge enrichment plants (GCEPs). While to date there has been no indication of declared, safeguarded GCEPs producing HEU, the massive separative work unit (SWU) processing power of modern GCEPs presents a significant latent risk of nuclear breakout and suggests the need for more timely detection of potential facility misuse. The Pacific Northwest National Laboratory is developing an unattended safeguards instrument, combining continuous aerosol particulate collection with uranium isotope assay, to provide timely HEU detection within a GCEP. This approach is based on laser vaporization of aerosol particulates, followed by laser spectroscopy to characterize the uranium enrichment level. We demonstrate enrichment assay, with relative isotope abundance uncertainty <5%, on individual micron-sized particles that are trace components within a mixture ‘background’ particles

Anheier, Norman C.; Bushaw, Bruce A.

2010-04-15T23:59:59.000Z

9

HEU age determination  

SciTech Connect

A criteria that a sample of highly enriched uranium (HEU) had come from a weapons stockpile and not newly produced in an enrichment plant is to show that the HEU had been produced a significant time in the past. The time since the HEU has produced in an enrichment plant is defined as the age of the HEU in this paper. The HEU age is determined by measuring quantitatively the daughter products {sup 230}Th and {sup 231}Pa of {sup 234}U and {sup 235}U, respectively, by first chemical separation of the thorium and protactinium and then conducting alpha spectrometry of the daughter products.

Moorthy, A.R.; Kato, W.Y.

1994-12-31T23:59:59.000Z

10

Portable NDA Equipment for Enrichment Measurements in the HEU Transparency Program  

SciTech Connect

The Highly Enriched Uranium (HEU) Transparency Program has used portable nondestructive assay (NDA) equipment to measure the {sup 235}U enrichment of material subject to the transparency agreement since 1997. The equipment is based on the 'enrichment meter' method and uses low-resolution sodium iodide (NaI(Tl)) detectors. Although systems using high-purity germanium (HPGe) detectors can produce more accurate results we have found that the results with NaI(Tl) detectors are quite adequate for the requirements of the transparency agreement. This paper will describe the details of the equipment's operation, calibration, testing, and deployment in Russia. We will also provide a comparison of the units originally deployed in 1997 with the upgraded systems that were deployed in 2003.

Decman, D J; Bandong, B B; Wong, J L; Valentine, J D; Luke, S J

2008-06-02T23:59:59.000Z

11

HEU age determination  

SciTech Connect

A technique has been developed to determine the Highly Enriched Uranium (HEU) Age which is defined as the time since the HEU was produced in an enrichment process. The HEU age is determined from the ratios of relevant uranium parents and their daughters viz {sup 230}Th/{sup 234}U and {sup 231}Pa/{sup 235}U. Uranium isotopes are quantitatively measured by their characteristic gammas and their daughters by alpha spectroscopy. In some of the samples where HEU is enriched more than 99%, the only mode of HEU age determination is by the measurement of {sup 231}Pa since there is negligible quantity of {sup 230}Th due to very low atom concentrations of {sup 234}U in the sample. In this paper we have presented data and methodology of finding the age of two HEU samples.

Moorthy, A.R.; Kato, W.Y.

1995-08-01T23:59:59.000Z

12

Detection of illicit HEU production in gaseous centrifuge enrichment plants using neutron counting techniques on product cylinders  

SciTech Connect

Innovative and novel safeguards approaches are needed for nuclear energy to meet global energy needs without the threat of nuclear weapons proliferation. Part of these efforts will include creating verification techniques that can monitor uranium enrichment facilities for illicit production of highly-enriched uranium (HEU). Passive nondestructive assay (NDA) techniques will be critical in preventing illicit HEU production because NDA offers the possibility of continuous and unattended monitoring capabilities with limited impact on facility operations. Gaseous centrifuge enrichment plants (GCEP) are commonly used to produce low-enriched uranium (LEU) for reactor fuel. In a GCEP, gaseous UF{sub 6} spins at high velocities in centrifuges to separate the molecules containing {sup 238}U from those containing the lighter {sup 235}U. Unfortunately, the process for creating LEU is inherently the same as HEU, creating a proliferation concern. Insuring that GCEPs are producing declared enrichments poses many difficult challenges. In a GCEP, large cascade halls operating thousands of centrifuges work together to enrich the uranium which makes effective monitoring of the cascade hall economically prohibitive and invasive to plant operations. However, the enriched uranium exiting the cascade hall fills product cylinders where the UF{sub 6} gas sublimes and condenses for easier storage and transportation. These product cylinders hold large quantities of enriched uranium, offering a strong signal for NDA measurement. Neutrons have a large penetrability through materials making their use advantageous compared to gamma techniques where the signal is easily attenuated. One proposed technique for detecting HEU production in a GCEP is using neutron coincidence counting at the product cylinder take off stations. This paper discusses findings from Monte Carlo N-Particle eXtended (MCNPX) code simulations that examine the feasibility of such a detector.

Freeman, Corey R [Los Alamos National Laboratory; Geist, William H [Los Alamos National Laboratory

2010-01-01T23:59:59.000Z

13

Unattended Monitoring of HEU Production in Gaseous Centrifuge Enrichment Plants using Automated Aerosol Collection and Laser-based Enrichment Assay  

Science Conference Proceedings (OSTI)

Nuclear power is enjoying rapid growth as government energy policies and public demand shift toward low carbon energy production. Pivotal to the global nuclear power renaissance is the development and deployment of robust safeguards instrumentation that allows the limited resources of the IAEA to keep pace with the expansion of the nuclear fuel cycle. Undeclared production of highly enriched uranium (HEU) remains a primary proliferation concern for modern gaseous centrifuge enrichment plants (GCEPs), due to their massive separative work unit (SWU) processing power and comparably short cascade equilibrium timescale. The Pacific Northwest National Laboratory is developing an unattended safeguards instrument, combining continuous aerosol particulate collection with uranium isotope assay, to provide timely detection of HEU production within a GCEP. This approach is based on laser vaporization of aerosol particulates, followed by laser spectroscopy to characterize the uranium enrichment level. Our prior investigation demonstrated single-shot detection sensitivity approaching the femtogram range and relative isotope ratio uncertainty better than 10% using gadolinium as a surrogate for uranium. In this paper we present measurement results on standard samples containing traces of depleted, natural, and low enriched uranium, as well as measurements on aerodynamic size uranium particles mixed in background materials (e.g., dust, minerals, soils). Improvements and optimizations in the detection electronics, signal timing, calibration, and laser alignment have lead to significant improvements in detection sensitivity and enrichment accuracy, contributing to an overall reduction in the false alarm probability. The sample substrate media was also found to play a significant role in facilitating laser-induced vaporization and the production of energetic plasma conditions, resulting in ablation optimization and further improvements in the isotope abundance sensitivity.

Anheier, Norman C.; Bushaw, Bruce A.

2010-08-11T23:59:59.000Z

14

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

SciTech Connect

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

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

1995-03-01T23:59:59.000Z

15

Portable NDA equipment for enrichment measurements for the HEU transparency program  

SciTech Connect

In October 1996, the Department of Energy (DOE) and MINATOM agreed to use portable non-destructive assay (NDA) equipment to measure the {sup 235}U enrichment of material subject to the HEU Transparency agreement. A system based on the ''enrichment meter'' method and high-purity germanium (HPGe) detectors had been previously developed for this application. Instead, sodium iodide (NaI) detectors were chosen to measure {sup 235}U enrichment because HPGe systems might reveal sensitive information. Although the accuracy of the NaI systems is less than an HPGe system, it still satisfies the transparency requirements. The equipment consists of a collimated NaI detector, a Canberra Inspector Multi-channel Analyzer, and a laptop computer. The units have been used to confirm the enrichment of material at Russian facilities since January 1997. This paper compares the performance of the NaI systems with the HPGe system and discusses some significant differences.

Decman, D J; Glaser, J; Hernandez, J M; Luke, S J

1999-07-20T23:59:59.000Z

16

Active interrogation of highly enriched uranium  

SciTech Connect

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

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

2004-01-01T23:59:59.000Z

17

Vietnam HEU Removal | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

NNSANews posted a photo: Vietnam HEU Removal A convoy escorting the last highly enriched uranium in Vietnam departs Dalat. Facebook Twitter Youtube Flickr Headlines Jul 23,...

18

Vietnam HEU Removal | National Nuclear Security Administration  

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

Removal NNSANews posted a photo: Vietnam HEU Removal A truck carrying the last highly enriched uranium in Vietnam winds through the Vietnamese countryside. Facebook Twitter Youtube...

19

The U. S. -Russian HEU Agreement: Internal safeguards to prevent diversion of HEU  

SciTech Connect

Under the U.S.-Russian HEU agreement, approximately 500 tons of highly enriched uranium (HEU) from large-scale dismantlement of former Soviet nuclear warheads will be transformed into products not usable in nuclear weapons. According to the agreement, Russian facilities will convert and blend down HEU to low-enriched uranium (LEU) fuel for nuclear reactors. However, HEU is vulnerable to insider diversion during processing operations. The paper describes the principal HEU diversion vulnerabilities at the plant, and recommends a strong internal preventive safeguards system. 27 refs., 2 figs., 1 tab.

Bukharin, O.; Hunt, H.M. (Princeton Univ., NJ (United States))

1994-01-01T23:59:59.000Z

20

Disposition of excess highly enriched uranium status and update  

SciTech Connect

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

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

1997-09-01T23:59:59.000Z

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


21

Removal of Last Remaining HEU from Vietnam - Time Lapse Video...  

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

(MOST) and the Russian Federation successfully removed 11 kilograms of highly enriched uranium (HEU) from the Dalat Nuclear Research Institute. This is the eleventh country...

22

Transportable calorimeter measurements of highly enriched uranium  

SciTech Connect

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

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

1997-11-01T23:59:59.000Z

23

Isotope Enrichment Detection by Laser Ablation - Laser Absorption Spectrometry: Automated Environmental Sampling and Laser-Based Analysis for HEU Detection  

SciTech Connect

The global expansion of nuclear power, and consequently the uranium enrichment industry, requires the development of new safeguards technology to mitigate proliferation risks. Current enrichment monitoring instruments exist that provide only yes/no detection of highly enriched uranium (HEU) production. More accurate accountancy measurements are typically restricted to gamma-ray and weight measurements taken in cylinder storage yards. Analysis of environmental and cylinder content samples have much higher effectiveness, but this approach requires onsite sampling, shipping, and time-consuming laboratory analysis and reporting. Given that large modern gaseous centrifuge enrichment plants (GCEPs) can quickly produce a significant quantity (SQ ) of HEU, these limitations in verification suggest the need for more timely detection of potential facility misuse. The Pacific Northwest National Laboratory (PNNL) is developing an unattended safeguards instrument concept, combining continuous aerosol particulate collection with uranium isotope assay, to provide timely analysis of enrichment levels within low enriched uranium facilities. This approach is based on laser vaporization of aerosol particulate samples, followed by wavelength tuned laser diode spectroscopy to characterize the uranium isotopic ratio through subtle differences in atomic absorption wavelengths. Environmental sampling (ES) media from an integrated aerosol collector is introduced into a small, reduced pressure chamber, where a focused pulsed laser vaporizes material from a 10 to 20-µm diameter spot of the surface of the sampling media. The plume of ejected material begins as high-temperature plasma that yields ions and atoms, as well as molecules and molecular ions. We concentrate on the plume of atomic vapor that remains after the plasma has expanded and then cooled by the surrounding cover gas. Tunable diode lasers are directed through this plume and each isotope is detected by monitoring absorbance signals on a shot-to-shot basis. The media is translated by a micron resolution scanning system, allowing the isotope analysis to cover the entire sample surface. We also report, to the best of our knowledge, the first demonstration of laser-based isotopic measurements on individual micron-sized particles that are minor target components in a much larger heterogeneous mix of ‘background’ particles. This composition is consistent with swipe and environmental aerosol samples typically collected for safeguards ES purposes. Single-shot detection sensitivity approaching the femtogram range and relative isotope abundance uncertainty better than 10% has been demonstrated using gadolinium isotopes as surrogate materials.

Anheier, Norman C.; Bushaw, Bruce A.

2010-01-01T23:59:59.000Z

24

HEU age determination  

SciTech Connect

A new technique has been developed to determine the age of highly enriched uranium (HEU) in solids. Uranium age is defined as the time since the uranium-containing material was last subjected to a process capable of separating uranium from its radioactive-decay daughters. [Most chemical processing, uranium enrichment, volatilization processes, and phase transformations (especially relevant for uranium hexafluoride) can result in separation of the uranium parent material from the decay-product daughters.] Determination of the uranium age, as defined here, may be relevant in verifying arms-control agreements involving uranium-containing nuclear weapons. The HEU age is determined from the ratios of relevant uranium daughter isotopes and their parents, viz {sup 230}Th/{sup 234}U and {sup 231}Pa/{sup 235}U. Uranium isotopes are quantitatively measured by their characteristic gamma rays and their daughters by alpha spectroscopy. In some of the samples, where HEU is enriched more than 99%, the only mode of HEU age determination is by the measurement of {sup 231}Pa since there is negligible quantity of {sup 230}Th due to very low atom concentrations of {sup 234}U in the samples. In this report the methodology and the data for determining the age of two HEU samples are presented.

Moorthy, A.R.; Kato, W.Y.

1997-07-01T23:59:59.000Z

25

Evaluation of the thermal-hydraulic operating limits of the HEU-LEU transition cores for the MIT Research Reactor.  

E-Print Network (OSTI)

??The MIT Research Reactor (MITR) is in the process of conducting a design study to convert from High Enrichment Uranium (HEU) fuel to Low Enrichment… (more)

Wang, Yunzhi (Yunzhi Diana)

2009-01-01T23:59:59.000Z

26

Accelerating the Reduction of Excess Russian Highly Enriched Uranium  

SciTech Connect

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

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

2004-02-18T23:59:59.000Z

27

EIS-0240: Disposition of Surplus Highly Enriched Uranium  

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

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

28

U.S. Transparency monitoring under the U.S./Russian HEU purchase agreement  

SciTech Connect

The conversion of Highly Enriched Uranium (HEU) metal to low enriched uranium (LEU) takes place at four Russian sites. HEU metal to oxide processing began in 1994 with shipments of HEU oxide from the Siberian Chemical Enterprise (SChE) to the Ural Electrochemical Integrated Plant (UEIP) fluorination and blending facility. U.S. transparency monitoring at these facilities began in February 1996. In 1996, fluorination and blending operations began at the Electrochemical Plant (ECP). In 1997, additional HEU metal to oxide was added at the Mayak Production Association (MPA), and additional fluorination and blending operations have been performed at SChE. U.S. transparency monitoring at these facilities is intended to provide confidence that HEU weapons components are received, that the HEU metal is converted to HEU oxide, and that the HEU is blended to LEU prior to shipment to the U.S. Enrichment Corporation (USEC). The monitoring begins with observation of HEU weapon components in sealed containers, including confirmation of the {sup 235}U enrichment using U.S. nondestructive assay (NDA) equipment. The feeding of HEU metal shavings to the oxidation process and the subsequent packaging of the HEU oxide for shipment to the fluorination and blending facilities are then monitored. At those facilities, monitors are allowed to witness the fluorination and blending of the HEU into LEU. Monitors are allowed to use the NDA instrumentation to confirm that HEU is being processed. A series of process and material accountancy documents are provided to U.S. monitors.

Benton, J; Dougherth, D R; Glaser, J W; Thomas, D C

1999-07-27T23:59:59.000Z

29

Photon and neutron active interrogation of highly enriched uranium.  

SciTech Connect

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

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

2004-01-01T23:59:59.000Z

30

Passive Time Coincidence Measurements with HEU Oxide Fuel Pins  

SciTech Connect

Passive time coincidence measurements have been performed on highly enriched uranium (HEU) oxide fuel pins at the Idaho National Laboratory Power Burst Facility. These experiments evaluate HEU detection capability using passive coincidence counting when utilizing moderated 3He tubes. Data acquisition was performed with the Nuclear Material Identification System (NMIS) to calculate the neutron coincidence time distributions. The amounts of HEU measured were 1 kg, 4 kg, and 8 kg in sealed 55-gallon drums. Data collected with the 3He tubes also include passive measurement of 31 kg of depleted uranium (DU) in order to determine the ability to distinguish HEU from DU. This paper presents results from the measurements.

McConchie, Seth M [ORNL; Hausladen, Paul [ORNL; Mihalczo, John T [ORNL

2008-01-01T23:59:59.000Z

31

Comparison of HEU and LEU neutron spectra in irradiation facilities at the Oregon State TRIGA® reactor.  

E-Print Network (OSTI)

??In 2008, the Oregon State TRIGA® Reactor (OSTR) was converted from highly-enriched uranium (HEU) fuel lifetime improvement plan (FLIP) fuel to low-enriched uranium (LEU) fuel.… (more)

[No author

2012-01-01T23:59:59.000Z

32

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

E-Print Network (OSTI)

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

Newton, Thomas Henderson

2006-01-01T23:59:59.000Z

33

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

E-Print Network (OSTI)

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

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

2008-01-01T23:59:59.000Z

34

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

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

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

35

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

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

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

36

Unallocated Off-Specification Highly Enriched Uranium: Recommendations for Disposition  

Science Conference Proceedings (OSTI)

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

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

2002-02-27T23:59:59.000Z

37

Surplus U.S. Highly Enriched Uranium (HEU) Disposition | National...  

National Nuclear Security Administration (NNSA)

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

38

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

SciTech Connect

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

Snider, J.D.

1996-02-01T23:59:59.000Z

39

Evaluation of the thermal-hydraulic operating limits of the HEU-LEU transition cores for the MIT Research Reactor  

E-Print Network (OSTI)

The MIT Research Reactor (MITR) is in the process of conducting a design study to convert from High Enrichment Uranium (HEU) fuel to Low Enrichment Uranium (LEU) fuel. The currently selected LEU fuel design contains 18 ...

Wang, Yunzhi (Yunzhi Diana)

2009-01-01T23:59:59.000Z

40

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

SciTech Connect

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

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

2008-06-05T23:59:59.000Z

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


41

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

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

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

42

HEU Transparency Implementation Program and its Radiation Safety Program  

SciTech Connect

In February 1993, the Governments of the United States (U.S.) and the Russian Federation (R.F.) signed a bilateral Agreement for the U.S. purchase of low enriched uranium (LEU) derived from 500 metric tons (MT) of highly enriched uranium (HEU) resulting from the dismantlement of Russian nuclear weapons. The HEU Purchase Agreement serves important national security and nonproliferation policy imperatives for both countries since its implementation reduces the quantity of surplus Russian HEU that could be stolen and diverted for weapons use. In return, Russia receives much needed U.S. dollars over a 20-year delivery period. In 2001, Russia received over half a billion US dollars from the purchase of the LEU blended from 30 MT HEU. As part of this Agreement, transparency rights were agreed upon that provide confidence to both governments that the nonproliferation objectives of the Agreement are being fulfilled. While the U.S. Department of State, in concert with the U.S. Department of Energy's (DOE) National Nuclear Security Administration (NNSA) is responsible negotiating transparency rights associated with this nuclear material, the NNSA is responsible for implementing those rights. These rights allow U.S. and R.F., personnel (called ''monitors'') to visit the processing facilities and observe the steps for processing the HEU into fuel for nuclear reactors. In this fashion, the processing of HEU to LEU is made ''transparent.'' For DOE, there are three transparency objectives: (1) that the HEU is extracted from nuclear weapons, (2) that this same HEU is oxidized, and (3) that the HEU is blended into LEU. For MINATOM, the transparency objective is: (1) that the LEU is fabricated into fuel for commercial nuclear power reactors: The transparency is based on visits by designated transparency monitors (100 preapproved U.S. and Russian monitors) with specific rights to monitor and to access storage and processing areas to provide confidence that the nonproliferation goals of the agreement are met. The Highly Enriched Uranium (HEU) Transparency Implementation Program (TIP), within NNSA implements the transparency provisions of the bilateral agreement. It is constantly making progress towards meeting its objectives and gathering the information necessary to confirm that Russian weapons-usable HEU is being blended into LEU. Since the first shipment in 1995 through December 2001, a total of 141 MT of weapons-grade HEU, about 28% of the agreed total and equivalent to 5,650 nuclear weapons, was converted to LEU, further reducing the threat of this material returning back into nuclear weapons. In the year 2001, the LEU sold to electric utility customers for fuel was sufficient to supply the annual fuel needs for about 50 percent of the U.S. installed nuclear electrical power generation capacity. There are four primary uranium processing activities involved in converting HEU metal components extracted from dismantled nuclear weapons into fuel for power reactors: (1) Converting HEU metal to purified HEU oxide; (2) Converting purified HEU oxide to HEU hexafluoride; (3) Downblending HEU hexafluoride to LEU hexafluoride; and (4) Converting LEU hexafluoride into reactor fuel. The first three processes are currently being performed at four Russian nuclear processing facilities: Mayak Production Association (MPA), Electrochemical Plant (ECP), Siberian Chemical Enterprise (SChE), and Ural Electrochemical Integrated Plant (UEIP). Following the blending down of HEU, the LEU hexafluoride is loaded into industry, standard 30B cylinders at the downblending facilities and transported to St. Petersburg, Russia. From there the LEU is shipped by sea to the United States where it is converted into fuel to be used in nuclear power plants. There are six U.S. facilities processing LEU subject to the HEU purchase agreement: the Portsmouth uranium enrichment plant, Global Nuclear Fuel -America, Framatome-Lynchburg, Framatome-Richland, Westinghouse-Hematite, and Westinghouse Fuel Fabrication Facility.

Radev, R

2002-01-31T23:59:59.000Z

44

Blenddown Monitoring System for HEU transparency  

SciTech Connect

The High Enriched Uranium (HEU) Purchase Agreement between the US and the Russian Federation (RF) provides for the monitoring of the blending of highly enriched uranium (500 metric tons) with low enrichment blend stock uranium (LEU) to produce commercial reactor-grade material for use in US reactors. A Blend Down Monitoring System (BDMS) has been developed by the US Department of Energy (DOE) to provide unattended monitoring of the HEU blending operations at the Russian facilities. It is configured to monitor the mass flow rate developed by the Oak Ridge National Laboratory (ORNL) and {sup 235}U isotopic enrichment developed by Los Alamos National Laboratory (LANL) of gaseous UF{sub 6} in three separate flow streams at a blending tee.

Mihalczo, J.T.

2000-02-01T23:59:59.000Z

45

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

SciTech Connect

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

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

2010-03-01T23:59:59.000Z

46

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

SciTech Connect

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

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

1997-03-28T23:59:59.000Z

47

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

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

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

48

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

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

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

49

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

SciTech Connect

This Conversion and Blending Facility (CBF) will have two missions: (1) convert HEU materials into pure HEU oxide and (2) blend the pure HEU oxide with depleted and natural uranium oxide to produce an LWR grade LEU product. The primary emphasis of this blending operation will be to destroy the weapons capability of large, surplus stockpiles of HEU. The blended LEU product can only be made weapons capable again by the uranium enrichment process. To the extent practical, the chemical and isotopic concentrations of blended LEU product will be held within the specifications required for LWR fuel. Such blended LEU product will be offered to the United States Enrichment Corporation (USEC) to be sold as feed material to the commercial nuclear industry. Otherwise, blended LEU will be produced as a waste suitable for storage or disposal.

1995-07-05T23:59:59.000Z

50

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

SciTech Connect

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

1995-07-05T23:59:59.000Z

51

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

SciTech Connect

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

Christopher Landers; Igor Bolshinsky; Ken Allen; Stanley Moses

2010-07-01T23:59:59.000Z

52

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

SciTech Connect

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

1995-07-05T23:59:59.000Z

53

Highly Enriched Uranium Transparency Program  

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

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

54

Feasibility analyses for HEU to LEU fuel conversion of the LAUE Langivin Institute (ILL) High Flux Reactor (RHF).  

SciTech Connect

The High Flux Reactor (RHF) of the Laue Langevin Institute (ILL) based in Grenoble, France is a research reactor designed primarily for neutron beam experiments for fundamental science. It delivers one of the most intense neutron fluxes worldwide, with an unperturbed thermal neutron flux of 1.5 x 10{sup 15} n/cm{sup 2}/s in its reflector. The reactor has been conceived to operate at a nuclear power of 57 MW but currently operates at 52 MW. The reactor currently uses a Highly Enriched Uranium (HEU) fuel. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context, most worldwide research and test reactors have already started a program of conversion to the use of Low Enriched Uranium (LEU) fuel. A new type of LEU fuel based on a mixture of uranium and molybdenum (UMo) is expected to allow the conversion of compact high performance reactors like the RHF. This report presents the results of reactor design, performance and steady state safety analyses for conversion of the RHF from the use of HEU fuel to the use of UMo LEU fuel. The objective of this work was to show that is feasible, under a set of manufacturing assumptions, to design a new RHF fuel element that could safely replace the HEU element currently used. The new proposed design has been developed to maximize performance, minimize changes and preserve strong safety margins. Neutronics and thermal-hydraulics models of the RHF have been developed and qualified by benchmark against experiments and/or against other codes and models. The models developed were then used to evaluate the RHF performance if LEU UMo were to replace the current HEU fuel 'meat' without any geometric change to the fuel plates. Results of these direct replacement analyses have shown a significant degradation of the RHF performance, in terms of both neutron flux and cycle length. Consequently, ANL and ILL have collaborated to investigate alternative designs. A promising candidate design has been selected and studied, increasing the total amount of fuel without changing the external plate dimensions by relocating the burnable poison. In this way, changes required in the fuel element are reasonably small. With this new design, neutronics analyses have shown that performance could be maintained at a high level: 2 day decrease of cycle length (to 47.5 days at 58.3 MW) and 1-2% decrease of brightness in the cold and hot sources in comparison to the current typical operation. In addition, studies have shown that the thermal-hydraulic and shutdown margins for the proposed LEU design would satisfy technical specifications.

Stevens, J.; Tentner. A.; Bergeron, A.; Nuclear Engineering Division

2010-08-19T23:59:59.000Z

55

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

SciTech Connect

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

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

2012-06-07T23:59:59.000Z

56

HEU Transparency Program: Monitoring at the U.S. Permanent Presence Office in Russia  

SciTech Connect

In February 1993, the US and the Russian Federation signed an agreement that allows the US to purchase 500 tonnes of Russian low-enriched uranium (LEU) that has been blended down from the high-enriched uranium (HEU) from Russia's dismantled nuclear weapons. The agreement calls for the HEU to be blended down to LEU at Russian facilities and then shipped to the United states to be used for making commercial reactor fuel. This HEU Agreement was crafted to avoid the rigid verification procedures of previous arms control and nonproliferation treaties. In the United States, it is being implemented by the US Department of Energy (DOE) under the HEU Transparency Program. Transparency refers to agreed-upon measures intended to build confidence that the objectives of the HEU Agreement are being met. The objectives of the HEU Transparency Program are to ensure that (a) the HEU subject to the agreement is extracted from Russian nuclear weapons; (b) this same extracted HEU enters an oxidation facility and is oxidized therein; (c) the declared quantity of HEU is blended down to LEU; and (d) the LEU that is delivered to the United states, pursuant to the agreement, is fabricated into fuel for commercial nuclear reactors. The HEU Agreement gives Russian monitors access to the US Enrichment Corporation's Portsmouth Gaseous Diffusion Plant in Piketon, Ohio, and to the five US fuel fabrication facilities receiving the Russian uranium. In turn, US monitors have access to the four principal Russian plants that convert HEU to LEU. Currently, monitoring at three Russian facilities--the Mayak Production Association near Ozersk, Siberian Chemical Enterprise (SChE) near Tomsk, and Electrochemical Integrated Plant (ECP) near Zelenogorsk--is confined to periodic visits. However, US monitors have continuous access to the Ural Electrochemical Integrated Enterprise (UEIE) in Novouralsk through the US Permanent Presence Office (PPO) located there. This paper summarizes the monitoring activities and challenges involved in managing and coordinating the PPO.

Boggs, C.J.; Monette, F.A.; Hensley, J.E.

1999-07-01T23:59:59.000Z

57

U.S. transparency monitoring of HEU oxide conversion and blending to LEU hexafluoride at three Russian blending plants  

SciTech Connect

The down-blending of Russian highly enriched uranium (HEU) takes place at three Russian gaseous centrifuge enrichment plants. The fluorination of HEU oxide and down-blending of HEU hexafluoride began in 1994, and shipments of low enriched uranium (LEU) hexafluoride product to the United States Enrichment Corporation (USEC) began in 1995 US transparency monitoring under the HEU Purchase Agreement began in 1996 and includes a permanent monitoring presence US transparency monitoring at these facilities is intended to provide confidence that HEU is received and down-blended to LEU for shipment to USEC The monitoring begins with observation of the receipt of HEU oxide shipments, including confirmation of enrichment using US nondestructive assay equipment The feeding of HEU oxide to the fluorination process and the withdrawal of HEU hexafluoride are monitored Monitoring is also conducted where the blending takes place and where shipping cylinders are filled with LEU product. A series of process and material accountancy documents are provided to US monitors.

Leich, D., LLNL

1998-07-27T23:59:59.000Z

58

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

SciTech Connect

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

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

2011-10-01T23:59:59.000Z

59

TRANSPARENCY: Tracking Uranium under the U.S. / Russian HEU Purchase Agreement  

SciTech Connect

By the end of August, 2005, the Russia Federation delivered to the United States (U.S.) more than 7,000 metric tons (MT) of low enriched uranium (LEU) containing approximately 46 million SWU and 75,000 MT of natural uranium. This uranium was blended down from weapons-grade (nominally enriched to 90% {sup 235}U) highly enriched uranium (HEU) under the 1993 HEU Purchase Agreement that provides for the blend down of 500 MT HEU into LEU for use as fuel in commercial nuclear reactors. The HEU Transparency Program, under the National Nuclear Security Administration (NNSA), monitored the conversion and blending of the more than 250 MT HEU used to produce this LEU. The HEU represents more than half of the 500 MT HEU scheduled to be blended down through the year 2013 and is equivalent to the elimination of more than 10,000 nuclear devices. The HEU Transparency Program has made considerable progress in its mission to develop and implement transparency measures necessary to assure that Russian HEU extracted from dismantled Russian nuclear weapons is blended down into LEU for delivery to the United States. U.S. monitor observations include the inventory of in process containers, observation of plant operations, nondestructive assay measurements to determine {sup 235}U enrichment, as well as the examination of Material Control and Accountability (MC&A) documents. During 2005, HEU Transparency Program personnel will conduct 24 Special Monitoring Visits (SMVs) to four Russian uranium processing plants, in addition to staffing a Transparency Monitoring Office (TMO) at one Russian site.

Benton, J B; Decman, D J; Leich, D A

2005-10-19T23:59:59.000Z

60

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

SciTech Connect

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

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

2004-03-09T23:59:59.000Z

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


61

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

SciTech Connect

This Conversion and Blending Facility (CBF) will have two missions: (1) convert HEU materials to pure HEU uranyl nitrate (UNH) and (2) blend pure HEU UNH with depleted and natural UNH to produce HEU UNH crystals. The primary emphasis of this blending operation will be to destroy the weapons capability of large, surplus stockpiles of HEU. The blended LEU product can only be made weapons capable again by the uranium enrichment process. To the extent practical, the chemical and isotopic concentrations of blended LEU product will be held within the specifications required for LWR fuel. Such blended LEU product will be offered to the United States Enrichment Corporation (USEC) to be sold as feed material to the commercial nuclear industry. Otherwise, blended LEU Will be produced as a waste suitable for storage or disposal.

1995-07-05T23:59:59.000Z

62

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

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

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

63

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

SciTech Connect

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

Glosup, J; Mount, M E

1999-07-01T23:59:59.000Z

64

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

SciTech Connect

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

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

2004-07-27T23:59:59.000Z

65

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

SciTech Connect

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

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

1994-02-01T23:59:59.000Z

66

2009 Annual Health Physics Report for the HEU Transparency Program  

SciTech Connect

During the 2009 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. LLNL also provided technical expertise related to BDMS radioactive sources and Russian radiation safety regulatory compliance. For the calendar year 2009, there were 159 person-trips that required dose monitoring of the U.S. monitors. Of the 159 person-trips, 149 person-trips were SMVs and 10 person-trips were Transparency Monitoring Office (TMO) trips. There were 4 monitoring visits by TMO monitors to facilities other than UEIE and 10 to UEIE itself. LLNL's Hazard Control Department laboratories provided the dosimetry services for the HEU Transparency monitors. In 2009, the HEU Transparency activities in Russia were conducted in a radiologically safe manner for the HEU Transparency monitors in accordance with the expectations of the HEU Transparency staff, NNSA and DOE. The HEU Transparency Program now has over fifteen years of successful experience in developing and providing health and safety support in meeting its technical objectives.

Radev, R

2010-04-14T23:59:59.000Z

67

2009 Annual Health Physics Report for the HEU Transparency Program  

SciTech Connect

During the 2009 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. LLNL also provided technical expertise related to BDMS radioactive sources and Russian radiation safety regulatory compliance. For the calendar year 2009, there were 159 person-trips that required dose monitoring of the U.S. monitors. Of the 159 person-trips, 149 person-trips were SMVs and 10 person-trips were Transparency Monitoring Office (TMO) trips. There were 4 monitoring visits by TMO monitors to facilities other than UEIE and 10 to UEIE itself. LLNL's Hazard Control Department laboratories provided the dosimetry services for the HEU Transparency monitors. In 2009, the HEU Transparency activities in Russia were conducted in a radiologically safe manner for the HEU Transparency monitors in accordance with the expectations of the HEU Transparency staff, NNSA and DOE. The HEU Transparency Program now has over fifteen years of successful experience in developing and providing health and safety support in meeting its technical objectives.

Radev, R

2010-04-14T23:59:59.000Z

68

2011 Annual Health Physics Report for the HEU transparency Program  

SciTech Connect

During the 2008 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. They also provided technical expertise related to BDMS radioactive sources and Russian radiation safety regulatory compliance. For the calendar year 2008, there were 158 person-trips that required dose monitoring of the U.S. monitors. Of the 158 person-trips, 148 person-trips were SMVs and 10 person-trips were Transparency Monitoring Office (TMO) trips. There were 6 monitoring visits by TMO monitors to facilities other than UEIE and 8 to UEIE itself. There were three monitoring visits (source changes) that were back-to-back with a total of 24 monitors. LLNL's Hazard Control Department laboratories provided the dosimetry services for the HEU Transparency monitors. In 2008, the HEU Transparency activities in Russia were conducted in a radiologically safe manner for the HEU Transparency monitors in accordance with the expectations of the HEU Transparency staff, NNSA and DOE. The HEU Transparency now has thirteen years of successful experience in developing and providing health and safety support in meeting its technical objectives.

Radev, R

2012-04-30T23:59:59.000Z

69

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

SciTech Connect

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

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

1993-01-01T23:59:59.000Z

70

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

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

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

71

A simple method for rapidly processing HEU from weapons returns  

SciTech Connect

A method based on the use of a high temperature fluidized bed for rapidly oxidizing, homogenizing and down-blending Highly Enriched Uranium (HEU) from dismantled nuclear weapons is presented. This technology directly addresses many of the most important issues that inhibit progress in international commerce in HEU; viz., transaction verification, materials accountability, transportation and environmental safety. The equipment used to carry out the oxidation and blending is simple, inexpensive and highly portable. Mobile facilities to be used for point-of-sale blending and analysis of the product material are presented along with a phased implementation plan that addresses the conversion of HEU derived from domestic weapons and related waste streams as well as material from possible foreign sources such as South Africa or the former Soviet Union.

McLean, W. II; Miller, P.E.

1994-01-01T23:59:59.000Z

72

Development of a low enrichment uranium core for the MIT reactor  

E-Print Network (OSTI)

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

Newton, Thomas Henderson

2006-01-01T23:59:59.000Z

73

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

E-Print Network (OSTI)

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

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

2008-01-01T23:59:59.000Z

74

U.S. HEU Disposition Program | National Nuclear Security Administratio...  

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

47 MT of surplus HEU was transferred to the United States Enrichment Corporation (USEC) and was down-blended into LEU commercial nuclear reactor fuel (project completed June...

75

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

SciTech Connect

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

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

1985-02-01T23:59:59.000Z

76

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

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

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

77

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

Science Conference Proceedings (OSTI)

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

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

2010-02-01T23:59:59.000Z

78

Preliminary neutronics calculations for conversion of the Tehran research reactor core from HEU to LEU fuel  

SciTech Connect

The 5-MW highly enriched uranium (HEU)-fueled Tehran Research Reactor is considered for conversion to high-density, low-enriched uranium (LEU) fuel. A preliminary neutronics calculation is performed as part of the conversion goal. In this study, two cores are considered: the HEU reference core and a proposed LEU core similar to the reference core, and a proposed LEU core similar to the reference core, using standardized U[sub 3]Si[sub 2] plates with the option of different [sup 235]U loadings. Various possibilities are investigated for the conversion of HEU to LEU fuel elements with 20% enriched [sup 235]U loadings of 207 to 297 g [sup 235]U/element. For the same equilibrium cycle length, the fuels are compared for flux, power distribution, burnup, and reactivity.

Nejat, S.M.R. (McMaster Univ., Hamilton, Ontario (Canada). Dept. of Engineering Physics.)

1993-08-01T23:59:59.000Z

79

Highly Enriched Uranium Transparency Program | National Nuclear...  

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

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

80

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

E-Print Network (OSTI)

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

Plumer, Kevin E. (Kevin Edward)

2011-01-01T23:59:59.000Z

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


81

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

E-Print Network (OSTI)

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

Chiang, Keng-Yen

2012-01-01T23:59:59.000Z

82

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

SciTech Connect

In July 2008, the Global Threat Reduction Initiative and the IRT 2000 research reactor in Sofia, Bulgaria, operated by the Institute for Nuclear Research and Nuclear Energy (INRNE), safely shipped 6.4 kilograms of Russian origin highly enriched uranium (HEU) spent nuclear fuel (SNF) to the Russian Federation. The shipment, which resulted in the removal of all HEU from Bulgaria, was conducted by truck, barge, and rail modes of transport across two transit countries before reaching the final destination at the Production Association Mayak facility in Chelyabinsk, Russia. This paper describes the work, equipment, organizations, and approvals that were required to complete the spent fuel shipment and provides lessons learned that might assist other research reactor operators with their own spent nuclear fuel shipments.

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

2009-07-01T23:59:59.000Z

83

Estimation of the Performance of Multiple Active Neutron Interrogation Signatures for Detecting Shielded HEU  

SciTech Connect

A comprehensive modeling study has been carried out to evaluate the utility of multiple active neutron interrogation signatures for detecting shielded highly enriched uranium (HEU). The modeling effort focused on varying HEU masses from 1 kg to 20 kg; varying types of shields including wood, steel, cement, polyethylene, and borated polyethylene; varying depths of the HEU in the shields, and varying engineered shields immediately surrounding the HEU including steel, tungsten, and cadmium. Neutron and gamma-ray signatures were the focus of the study and false negative detection probabilities versus measurement time were used as a performance metric. To facilitate comparisons among different approaches an automated method was developed to generate receiver operating characteristic (ROC) curves for different sets of model variables for multiple background count rate conditions. This paper summarizes results or the analysis, including laboratory benchmark comparisons between simulations and experiments. The important impact engineered shields can play towards degrading detectability and methods for mitigating this will be discussed.

David L. Chichester; Scott J. Thompson; Scott M. Watson; James T. Johnson; Edward H. Seabury

2012-10-01T23:59:59.000Z

84

HEU to LEU Conversion and Blending Facility: UNH blending alternative to produce LEU UNH for commercial use  

SciTech Connect

US DOE is examining options for disposing of surplus weapons-usable fissile materials and storage of all weapons-usable fissile materials. The nuclear material is converted to a form that is more proliferation-resistant than the original form. Examining options for increasing the proliferation resistance of highly enriched uranium (HEU) is part of this effort. Five technologies for blending HEU will be assessed. This document provides data to be used in the environmental impact analysis for the UNH blending HEU disposition option. Process requirements, resource needs, employment needs, waste/emissions from plant, hazards, accident scenarios, and intersite transportation are discussed.

1995-09-01T23:59:59.000Z

85

Reflected Neutron Effects in Multiplicity Measurements of Bare HEU Assemblies  

SciTech Connect

In a passive multiplicity characterization of highly enriched uranium (HEU) assemblies, fission chains are initiated by the characteristically fast neutrons from spontaneous fission of {sup 238}U and {sup 235}U as well as cosmic-ray spallation neutrons. Active interrogation of HEU uses other physical mechanisms for starting chains by inducing fission from high-energy neutrons, high-energy gamma-rays, delayed neutrons, or thermal neutrons. In all cases a contribution to the initiation of fission chains is the reflection of neutrons that initially escape the assembly and re-enter it after undergoing some scattering. The reflected neutron flux is geometry dependent and a combination of fast and thermal energies. The reflected thermal neutron contribution occurs hundreds of microseconds after the beginning of the fission chain and can be distinguished from the cosmic-ray spallation neutrons unrelated to fission chains, resulting in an HEU detection signature with high signal-to-noise. However, the reflected thermal neutron flux can be eliminated with an efficient thermal neutron absorber to investigate reflected neutron effects. In this paper, active and passive multiplicity measurements with HEU oxide assemblies of up to 16 kg of fuel pins and HEU metal assemblies of up to five 18 kg storage castings are reported. Each case demonstrates the differences in HEU signature when a borated thermal neutron absorber is present and shows the various detectable signatures with 3He proportional counters, the standard detector for differential die-way and neutron multiplicity measurements, and liquid scintillators, a detector capable of operating on the timescale of fission chains.

McConchie, Seth M [ORNL; Hausladen, Paul [ORNL; Mihalczo, John T [ORNL

2010-01-01T23:59:59.000Z

86

Validation of a Monte Carlo based depletion methodology via High Flux Isotope Reactor HEU post-irradiation examination measurements  

Science Conference Proceedings (OSTI)

The purpose of this study is to validate a Monte Carlo based depletion methodology by comparing calculated post-irradiation uranium isotopic compositions in the fuel elements of the High Flux Isotope Reactor (HFIR) core to values measured using uranium mass-spectrographic analysis. Three fuel plates were analyzed: two from the outer fuel element (OFE) and one from the inner fuel element (IFE). Fuel plates O-111-8, O-350-1, and I-417-24 from outer fuel elements 5-O and 21-O and inner fuel element 49-I, respectively, were selected for examination. Fuel elements 5-O, 21-O, and 49-1 were loaded into HFIR during cycles 4, 16, and 35, respectively (mid to late 1960s). Approximately one year after each of these elements were irradiated, they were transferred to the High Radiation Level Examination Laboratory (HRLEL) where samples from these fuel plates were sectioned and examined via uranium mass-spectrographic analysis. The isotopic composition of each of the samples was used to determine the atomic percent of the uranium isotopes. A Monte Carlo based depletion computer program, ALEPH, which couples the MCNP and ORIGEN codes, was utilized to calculate the nuclide inventory at the end-of-cycle (EOC). A current ALEPH/MCNP input for HFIR fuel cycle 400 was modified to replicate cycles 4, 16, and 35. The control element withdrawal curves and flux trap loadings were revised, as well as the radial zone boundaries and nuclide concentrations in the MCNP model. The calculated EOC uranium isotopic compositions for the analyzed plates were found to be in good agreement with measurements, which reveals that ALEPH/MCNP can accurately calculate burn-up dependent uranium isotopic concentrations for the HFIR core. The spatial power distribution in HFIR changes significantly as irradiation time increases due to control element movement. Accurate calculation of the end-of-life uranium isotopic inventory is a good indicator that the power distribution variation as a function of space and time is accurately calculated, i.e. an integral check. Hence, the time dependent heat generation source terms needed for reactor core thermal hydraulic analysis, if derived from this methodology, have been shown to be accurate for highly enriched uranium (HEU) fuel.

Chandler, David [ORNL; Maldonado, G Ivan [ORNL; Primm, Trent [ORNL

2010-01-01T23:59:59.000Z

87

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

Science Conference Proceedings (OSTI)

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

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

2011-02-01T23:59:59.000Z

88

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

Science Conference Proceedings (OSTI)

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

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

2010-02-01T23:59:59.000Z

89

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

Science Conference Proceedings (OSTI)

An engineering design study of the conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel is ongoing at Oak Ridge National Laboratory. The computational models developed during fiscal year 2010 to search for an LEU fuel design that would meet the requirements for the conversion and the results obtained with these models are documented and discussed in this report. Estimates of relevant reactor performance parameters for the LEU fuel core are presented and compared with the corresponding data for the currently operating HEU fuel core. The results obtained indicate that the LEU fuel design would maintain the current performance of the HFIR with respect to the neutron flux to the central target region, reflector, and beam tube locations under the assumption that the operating power for the reactor fueled with LEU can be increased from the current value of 85 MW to 100 MW.

Ilas, Germina [ORNL; Primm, Trent [ORNL

2011-05-01T23:59:59.000Z

90

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

Science Conference Proceedings (OSTI)

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

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

2009-03-01T23:59:59.000Z

91

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

SciTech Connect

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

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

2004-07-01T23:59:59.000Z

92

Measurements of the HEU and LEU in-core spectra at the Ford Nuclear Reactor  

SciTech Connect

The Ford Nuclear Reactor (FNR) at the University of Michigan has been serving as the test site for a low-enriched uranium (LEU) fuel whole-core demonstration. As part of the experimental program, the differential neutron spectrum has been measured in a high-enriched uranium (HEU) core and an LEU core. The HEU and LEU spectra were determined by unfolding the measured activities of foils that were irradiated in the reactor. When the HEU and LEU spectra are compared from 1 MeV to 10 MeV, significant differences between the two spectra are apparent below 10 eV. These are probably caused by the additional /sup 238/U resonance absorption in the LEU fuel. No measurable difference occurs in the shape of the spectra above 1 MeV. 7 refs., 6 figs., 2 tabs.

Wehe, D.K.; King, J.S.; Lee, J.C.; Martin, W.R.

1984-01-01T23:59:59.000Z

93

HEU to LEU conversion and blending facility: Metal blending alternative to produce LEU oxide for disposal  

SciTech Connect

US DOE is examining options for disposing of surplus weapons-usable fissile materials and storage of all weapons-usable fissile materials. The nuclear material is converted to a form more proliferation- resistant than the original form. Blending HEU (highly enriched uranium) with less-enriched uranium to form LEU has been proposed as a disposition option. Five technologies are being assessed for blending HEU. This document provides data to be used in environmental impact analysis for the HEU-LEU disposition option that uses metal blending with an oxide waste product. It is divided into: mission and assumptions, conversion and blending facility descriptions, process descriptions and requirements, resource needs, employment needs, waste and emissions from plant, hazards discussion, and intersite transportation.

1995-09-01T23:59:59.000Z

94

Overview of Russian HEU transparency issues  

SciTech Connect

The U.S. has signed an agreement with the Russian Federation for the purchase of 500 metric tons of highly-enriched uranium (HEU) taken from dismantled nuclear weapons. The HEU will be blended down to low-enriched uranium and will be transported to the U.S. to be used by fuel fabricators to make fuel for commercial nuclear power plants. Both the U.S. and Russia have been preparing to institute transparency measures to provide assurance that nonproliferation and arms control objectives specified in the agreement are met. This paper provides background information on the original agreement and on subsequent negotiations with the Russians, as well as discussion of technical aspects of developing transparency measures suited to the facilities and processes which are expected to be involved. Transparency has been defined as those agreed-upon measures which build confidence that arms control and non-proliferation objectives shared by the parties are met. Transparency is a departure from exhaustive, detailed arms control verification regimes of past agreements, which were based on a presumption of detecting transgressions as opposed to confirming compliance.

Kempf, C.R. [Brookhaven National Lab., Upton, NY (United States); Bieniawski, A. [USDOE Office of Arms Control and Nonproliferation, Washington, DC (United States)

1993-09-01T23:59:59.000Z

95

Validation of ATR Fission Power Deposition Fraction in HEU and LEU Fuel Plates  

SciTech Connect

The Advanced Test Reactor (ATR) is a high power (250 MW), high neutron flux research reactor operating in the United States. Powered with highly enriched uranium (HEU), the ATR has a maximum unperturbed thermal neutron flux rating of 1.0 x 1015 n/cm2–s. Because of its high power and large test volumes located in high flux areas, the ATR is an ideal candidate for assessing the feasibility of converting an HEU driven reactor to a low-enriched core. A detailed plate-by-plate MCNP ATR full core model has been developed and validated for the low-enriched uranium (LEU) fuel conversion feasibility study. Using this model, an analysis has been performed to determine the LEU density and U-235 enrichment required in the fuel meat to yield equivalent K-eff versus effective full power days (EFPDs) between the HEU and LEU cores. This model has also been used to optimize U-235 content of the LEU core, minimizing the differences in K-eff and heat flux profile between the HEU and LEU cores at 115 MW total core power for 125 EFPDs. The LEU core conversion feasibility study evaluated foil type (U-10Mo) fuel with the LEU reference design of 19.7 wt% U-235 enrichment. The LEU reference design has a fixed fuel meat thickness of 0.330 mm and can sustain the same operating cycle length as the HEU fuel. Heat flux and fission power density are parameters that are proportional to the fraction of fission power deposited in fuel. Thus, the accurate determination of the fraction of fission power deposited in the fuel is important to ATR nuclear safety. In this work, a new approach was developed and validated, the Tally Fuel Cells Only (TFCO) method. This method calculates and compares the fission power deposition fraction between HEU and LEU fuel plates. Due to the high density of the U-10Mo LEU fuel, the fission ?-energy deposition fraction is 37.12%, which is larger than the HEU’s ?-energy deposition fraction of 19.7%. As a result, the fuel decay heat cooling will need to be improved. During the power operation, the total fission energy (200 MeV per fission) deposition fraction of LEU and HEU are 90.9% and 89.1%, respectively.

G. S. Chang

2008-09-01T23:59:59.000Z

96

Highly Enriched Uranium Disposition | National Nuclear Security...  

National Nuclear Security Administration (NNSA)

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

97

HEU to LEU conversion and blending facility: UNH blending alternative to produce LEU oxide for disposal  

SciTech Connect

The United States Department of Energy (DOE) is examining options for the disposition of surplus weapons-usable fissile materials and storage of all weapons-usable fissile materials. Disposition is a process of use or disposal of material that results in the material being converted to a form that is substantially and inherently more proliferation-resistant than is the original form. Examining options for increasing the proliferation resistance of highly enriched uranium (HEU) is part of this effort. This report provides data to be used in the environmental impact analysis for the uranyl nitrate hexahydrate blending option to produce oxide for disposal. This the Conversion and Blending Facility (CBF) alternative will have two missions (1) convert HEU materials into HEU uranyl nitrate (UNH) and (2) blend the HEU uranyl nitrate with depleted and natural assay uranyl nitrate to produce an oxide that can be stored until an acceptable disposal approach is available. The primary emphasis of this blending operation will be to destroy the weapons capability of large, surplus stockpiles of HEU. The blended LEU product can only be made weapons capable again by the uranium enrichment process. The blended LEU will be produced as a waste suitable for storage or disposal.

1995-09-01T23:59:59.000Z

98

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

SciTech Connect

This report describes progress made during FY11 in ORNL activities to support converting the High Flux Isotope Reactor (HFIR) from high-enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. Conversion from HEU to LEU will require a change in fuel form from uranium oxide to a uranium-molybdenum (UMo) alloy. With both radial and axial contouring of the fuel foil and an increase in reactor power to 100 MW, calculations indicate that the HFIR can be operated with LEU fuel with no degradation in performance to users from the current levels achieved with HEU fuel. Studies are continuing to demonstrate that the fuel thermal safety margins can be preserved following conversion. Studies are also continuing to update other aspects of the reactor steady state operation and accident response for the effects of fuel conversion. Technical input has been provided to Oregon State University in support of their hydraulic testing program. The HFIR conversion schedule was revised and provided to the GTRI program. In addition to HFIR conversion activities, technical support was provided directly to the Fuel Fabrication Capability program manager.

Renfro, David G [ORNL; Cook, David Howard [ORNL; Freels, James D [ORNL; Griffin, Frederick P [ORNL; Ilas, Germina [ORNL; Sease, John D [ORNL; Chandler, David [ORNL

2012-03-01T23:59:59.000Z

99

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

SciTech Connect

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

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

1994-11-01T23:59:59.000Z

100

Design and Performance Considerations for the Quantitative Measurement of HEU Residues Resulting from 99Mo Production  

SciTech Connect

Molybdenum-99 is produced by the irradiation of high-enriched uranium (HEU) resulting in the accumulation of large quantities of HEU residues. In general, these residues are not recycled but are either disposed of or stored in containers with surface exposure rates as high as 100 R/h. The 235U content of these waste containers must be quantified for both accountability and waste disposal purposes. The challenges of quantifying such difficult-to-assay materials are discussed, along with performance estimates for each of several potential assay options. In particular, the design and performance of a High Activity Active Well Coincidence Counting (HA-AWCC) system designed and built specifically for these irradiated HEU waste materials are presented.

McElroy, Robert Dennis [ORNL; Chapman, Jeffrey Allen [ORNL; Bogard, James S [ORNL; Belian, Anthony P [Los Alamos National Laboratory (LANL)

2011-01-01T23:59:59.000Z

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101

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

SciTech Connect

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

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

2007-11-01T23:59:59.000Z

102

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

SciTech Connect

In partial response to a Department of Energy (DOE) request to evaluate the state of measurements of special nuclear material, Lawrence Livermore National Laboratory (LLNL) evaluated and classified all highly enriched uranium (HEU) oxide items in its inventory. Because of a lack of traceable HEU standards, no items were deemed to fit the category of well measured. A subsequent DOE-HQ sponsored survey by New Brunswick Laboratory resulted in their preparation of certified reference material (CRM) 149 [Uranium (93% Enriched) Oxide-U{sub 3}O{sub 8} Standard for Neutron Counting Measurements], a unit of which was delivered to LLNL in October of 1999. This paper describes the approach to calibration of the LLNL passive-active neutron drum (PAN) shuffler for measurement of poorly measured/unmeasured HEU oxide inventory. Included are discussions of (1) the calibration effort, including the development of the mass calibration curve; (2) the results from an axial and radial mapping of the detector response over a wide region of the PAN shuffler counting chamber, and (3) an error model for the total (systematic + random) uncertainty in the predicted mass that includes the uncertainties in calibration and sample position.

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

2000-06-16T23:59:59.000Z

103

Air Shipment of Highly Enriched Uranium Spent Nuclear Fuel from Romania  

SciTech Connect

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

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

2010-07-01T23:59:59.000Z

104

2004 Annual Health Physics Report for the HEU Transparency Program  

SciTech Connect

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 dosimetry services for the HEU-TIP monitors.

Radev, R

2005-04-01T23:59:59.000Z

105

2005 Annual Health Physics Report for HEU Transparency Program  

SciTech Connect

During the 2005 calendar year, LLNL provided health physics support for the Highly Enriched Uranium Transparency Program (HEU-TP) 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 2005, there were 161 person-trips that required dose monitoring of the U.S. monitors. Of the 161 person-trips, 149 person-trips were SMVs and 12 person-trips were Transparency Monitoring Office (TMO) trips. Additionally, there were 11 monitoring visits by TMO monitors to facilities other than UEIE and 3 to UEIE itself. There were two monitoring visits (source changes) that were back to back with 16 monitors. Each of these concurring visits were treated as single person-trips for dosimetry purposes. Counted individually, there were 191 individual person-visits in 2005. The LLNL Safety Laboratories Division provided the dosimetry services for the HEU-TP monitors.

Radev, R

2006-04-21T23:59:59.000Z

106

2007 Annual Health Physics Report for the HEU Transparency Program  

SciTech Connect

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 (source changes) that were back to back with 14 monitors. LLNL's Hazard Control Division laboratories provided the dosimetry services for the HEU Transparency monitors.

Radev, R

2008-04-09T23:59:59.000Z

107

DOE/EA-1471: Environmental Assessment for the Transportation of Highly Enriched Uranium from the Russian Federation to the Y-12 National Security Complex and Finding of No Significant Impact (January 2004)  

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

EA for the Transportation of Highly Enriched Uranium from the Russian Federation to the Y-12 National Security Complex EA for the Transportation of Highly Enriched Uranium from the Russian Federation to the Y-12 National Security Complex i FINDING OF NO SIGNIFICANT IMPACT FOR THE TRANSPORTATION OF HIGHLY ENRICHED URANIUM FROM THE RUSSIAN FEDERATION TO THE Y-12 NATIONAL SECURITY COMPLEX ISSUED BY: United States Department of Energy ACTION: Finding of No Significant Impact SUMMARY: The United States (U.S.) Department of Energy (DOE) proposes to transport highly enriched uranium (HEU) from Russia to a secure storage facility in Oak Ridge, TN. This proposed action would allow the United States and Russia to accelerate the disposition of excess nuclear weapons materials in the interest of promoting nuclear disarmament, strengthening nonproliferation, and combating terrorism. The HEU

108

Pulsed D-D Neutron Generator Measurements of HEU Oxide Fuel Pins  

SciTech Connect

Pulsed neutron interrogation measurements have been performed on highly enriched uranium (HEU) oxide fuel pins and depleted uranium (DU) metal using a D-D neutron generator (2 x 10{sup 6} neutrons-s{sup -1}) and moderated {sup 3}He tubes at the Idaho National Laboratory Power Burst Facility. These measurements demonstrate the ability to distinguish HEU from DU by coincidence counting using a pulsed source. The amount of HEU measured was 8 kg in a sealed 55-gallon drum compared to 31 kg of DU. Neutron events were counted during and after the pulse with the Nuclear Materials Identification System (NMIS) and used to calculate the neutron coincidence time distributions. Passive measurements were also performed for comparison with the pulsed measurements. This paper presents the neutron coincidence distribution and Feynman variance results from the measurements.

McConchie, Seth M [ORNL; Hausladen, Paul [ORNL; Mihalczo, John T [ORNL; Blackburn, Brandon [Idaho National Laboratory (INL); Chichester, David [Idaho National Laboratory (INL)

2009-01-01T23:59:59.000Z

109

Ion-induced gammas for photofission interrogation of HEU.  

Science Conference Proceedings (OSTI)

High-energy photons and neutrons can be used to actively interrogate for heavily shielded special nuclear material (SNM), such as HEU (highly enriched uranium), by detecting prompt and/or delayed induced fission signatures. In this work, we explore the underlying physics for a new type of photon source that generates high fluxes of mono-energetic gamma-rays from low-energy (ion accelerator experiments and radiation transport modeling. The data were used to assess gamma and neutron yields, background, and photofission-induced signal levels from several (p,{gamma}) target materials under consideration.

Doyle, Barney Lee (Sandia National Laboratories, Albuquerque, NM); Antolak, Arlyn J.; Morse, Daniel H.; Provencio, Paula Polyak (Sandia National Laboratories, Albuquerque, NM)

2006-03-01T23:59:59.000Z

110

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

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

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

111

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

SciTech Connect

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

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

2006-07-01T23:59:59.000Z

112

Comparison and validation of HEU and LEU modeling results to HEU experimental benchmark data for the Massachusetts Institute of Technology MITR reactor.  

Science Conference Proceedings (OSTI)

The Massachusetts Institute of Technology Reactor (MITR-II) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. It delivers a neutron flux comparable to current LWR power reactors in a compact 6 MW core using Highly Enriched Uranium (HEU) fuel. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context, most research and test reactors both domestic and international have started a program of conversion to the use of Low Enriched Uranium (LEU) fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (UMo) is expected to allow the conversion of U.S. domestic high performance reactors like the MITR-II reactor. Towards this goal, comparisons of MCNP5 Monte Carlo neutronic modeling results for HEU and LEU cores have been performed. Validation of the model has been based upon comparison to HEU experimental benchmark data for the MITR-II. The objective of this work was to demonstrate a model which could represent the experimental HEU data, and therefore could provide a basis to demonstrate LEU core performance. This report presents an overview of MITR-II model geometry and material definitions which have been verified, and updated as required during the course of validation to represent the specifications of the MITR-II reactor. Results of calculations are presented for comparisons to historical HEU start-up data from 1975-1976, and to other experimental benchmark data available for the MITR-II Reactor through 2009. This report also presents results of steady state neutronic analysis of an all-fresh LEU fueled core. Where possible, HEU and LEU calculations were performed for conditions equivalent to HEU experiments, which serves as a starting point for safety analyses for conversion of MITR-II from the use of HEU fuel to the use of UMo LEU fuel.

Newton, T. H.; Wilson, E. H; Bergeron, A.; Horelik, N.; Stevens, J. (Nuclear Engineering Division); (MIT Nuclear Reactor Lab.)

2011-03-02T23:59:59.000Z

113

Subcritical Measurments Multiple HEU Metal Castings  

SciTech Connect

Experiments with the standard annular highly enriched uranium (HEU) metal castings at the Y-12 National Security Complex were performed in which up to five castings ({approx}90 kg) were assembled in a tightly packed array with minimal spacing between castings. The fission chain multiplication process was initiated by a time-tagged {sup 252}Cf spontaneously fissioning neutron source or by time and directionally tagged neutrons from a small portable DT neutron generator. The prompt neutron time behavior was measured with plastic scintillation detectors sensitive to the fast neutron (>1 MeV) and gamma ray without distinction. These experiments were performed to provide data to benchmark methods for the calculation of the prompt neutron time behavior. Previous measurements with a single casting have been reported. This paper presents the experimental results for multiple castings.

Mihalczo, John T [ORNL; Archer, Daniel E [ORNL; Wright, Michael C [ORNL; Mullens, James Allen [ORNL

2008-01-01T23:59:59.000Z

114

Comparison of HEU and LEU Fuel Neutron Spectrum for ATR Fuel Element and ATR Flux-Trap Positions  

SciTech Connect

The Advanced Test Reactor (ATR) is a high power and high neutron flux research reactor operating in the United States. Powered with highly enriched uranium (HEU), the ATR has a maximum thermal power rating of 250 MWth. Because of the high total core power and high neutron flux, the ATR is an ideal candidate for assessing the feasibility of converting an HEU driven reactor to a low-enriched core. An optimized low-enriched uranium (LEU) (U-10Mo) core conversion case, which can meet the project requirements, has been selected. However, LEU contains a significant quantity of high density U-238 (80.3 wt.%), which will harden the neutron spectrum in the core region. Based on the reference ATR HEU and the optimized LEU full core plate-by-plate (PBP) models, the present work investigates and compares the neutron spectra differences in the fuel element (FE), Northeast flux trap (NEFT), Southeast flux trap (SEFT), and East flux trap (EFT) positions. A detailed PBP MCNP ATR core model was developed and validated for fuel cycle burnup comparison analysis. The current ATR core with HEU U 235 enrichment of 93.0wt.% was used as the reference model. Each HEU fuel element contains 19 fuel plates with a fuel meat thickness of 0.508 mm (20 mil). In this work, an optimized LEU (U-10Mo) core conversion case with a nominal fuel meat thickness of 0.330 mm (13 mil) and the U-235 enrichment of 19.7 wt.% was used to calculate the impact of the neutron spectrum in FE and FT positions. MCNP-calculated results show that the neutron spectrum in the LEU FE is slightly harder than in the HEU FE, as expected. However, when neutrons transport through water coolant and beryllium (Be), the neutrons are thermalized to an equilibrium neutron spectrum as a function of water volume fraction in the investigated FT positions. As a result, the neutron spectrum differences of the HEU and LEU in the NEFT, SEFT, and EFT are negligible. To demonstrate that the LEU core fuel cycle performance can meet the Updated Final Safety Analysis Report (UFSAR) safety requirements, additional studies will be necessary to evaluate and compare safety parameters such as void reactivity and Doppler coefficients, control components worth (outer shim control cylinders, safety rods and regulating rod), and shutdown margins between the HEU and LEU cores.

G. S. Chang

2008-10-01T23:59:59.000Z

115

Implementation of the United States/Russian HEU Agreement: Current Status and Prospects  

SciTech Connect

During Calendar Year (CY) 2002, the Russian Federation (R.F.) delivered low enriched uranium (LEU) from the conversion and processing of 30 metric tons (MT) of weapons-grade (90% {sup 235}U assay) uranium. Through July 2003, the Highly Enriched Uranium (HEU) Transparency Implementation Program (TIP) will have monitored the conversion of over 190 MT HEU into LEU. This total represents about 38 percent of the projected 500 MT HEU scheduled to be blended down through the year 2013 and is equivalent to the destruction of 7,600 nuclear devices. The National Nuclear Security Administration's (NNSA) HEU-TIP monitors the processing of this HEU at four Russian uranium-processing plants. During CY 2002, United States (U.S.) personnel monitored this process for a total of 194 monitor-weeks by staffing a Transparency Monitoring Office (TMO) located in Novouralsk, and through a series of five-day Special Monitoring Visits (SMV) to the four plants. U.S. monitor observations include the inventory of in-process containers, the observation of operations and non-destructive assay measurements (NDA) to determine {sup 235}U enrichment, as well as the examination and validation of Russian Material Control and Accountability (MC&A) documents. In addition, the U.S. designed Blend Down Monitoring System (BDMS) installed at the Ural Electrochemical Integrated Plant (UEIP) in January 1999 monitored all HEU blended at that facility, which is about 50 percent of the HEU blended into LEU during CY 2002. Recently we installed a BDMS at the Electrochemical Plant (ECP) in Zelenogorsk and plans are underway to install a BDMS at the Siberian Chemical Enterprise (SChE) in Seversk in late 2004. On a very positive note, interpersonal interactions between U.S. and Russian technical experts continues to expand and have proven to be an important element of the transparency regime. On the tenth anniversary of the HEU Purchase Agreement, the Ministry of the R.F. for Atomic Energy (Minatom) also saluted the successful implementation of the government-to-government program as ''an example of the effective realization of bilateral cooperation in real disarmament''. This paper describes the Program's monitoring efforts and achievements at the four Russian uranium processing plants, and will touch upon the issues of transparency and the natural uranium component activities.

Rutkowski, E

2003-11-19T23:59:59.000Z

116

Highly Enriched Uranyl Nitrate in Annular Tanks with Concrete Reflection: 1 x 3 Line Array of Nested Pairs of Tanks  

Science Conference Proceedings (OSTI)

A series of seven experiments were performed at the Rocky Flats Critical Mass Laboratory beginning in August, 1980 (References 1 and 2). Highly enriched uranyl nitrate solution was introduced into a 1-3 linear array of nested stainless steel annular tanks. The tanks were inside a concrete enclosure, with various moderator and absorber materials placed inside and/or between the tanks. These moderators and absorbers included boron-free concrete, borated concrete, borated plaster, and cadmium. Two configurations included placing bottles of highly enriched uranyl nitrate between tanks externally. Another experiment involved nested hemispheres of highly enriched uranium placed between tanks externally. These three configurations are not evaluated in this report. The experiments evaluated here are part of a series of experiments, one set of which is evaluated in HEU-SOL-THERM-033. The experiments in this and HEU-SOL-THERM-033 were performed similarly. They took place in the same room and used the same tanks, some of the same moderators and absorbers, some of the same reflector panels, and uranyl nitrate solution from the same location. There are probably additional similarities that existed that are not identified here. Thus, many of the descriptions in this report are either the same or similar to those in the HEU-SOL-THERM-033 report. Seventeen configurations (sixteen of which were critical) were performed during seven experiments; six of those experiments are evaluated here with thirteen configurations. Two configurations were identical, except for solution height, and were conducted to test repeatability. The solution heights were averaged and the two were evaluated as one configuration, which gives a total of twelve evaluated configurations. One of the seventeen configurations was subcritical. Of the twelve critical configurations evaluated, nine were judged as acceptable as benchmarks.

James Cleaver; John D. Bess; Nathan Devine; Fitz Trumble

2009-09-01T23:59:59.000Z

117

Physical inventory verification exercise for a highly enriched uranium fabrication facility  

SciTech Connect

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

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

1986-01-01T23:59:59.000Z

118

Analysis of HEU samples from the ULBA Metallurgical Plant  

SciTech Connect

In early March 1994, eight highly enriched uranium (HEU) samples were collected from materials stored at the Ulba Metallurgical Plant in Oskamen (Ust Kamenogorsk), Kazakhstan. While at the plant site, portions of four samples were dissolved and analyzed by mass spectrograph at the Ulba analytical laboratory by Ulba analysts. Three of these mass spectrograph solutions and the eight HEU samples were subsequently delivered to the Y-12 Plant for complete chemical and isotopic analyses. Chemical forms of the eight samples were uranium metal chips, U0{sub 2} powder, uranium/beryllium oxide powder, and uranium/beryllium alloy rods. All were declared by the Ulba plant to have a uranium assay of {approximately}90 wt % {sup 235}U. The uranium/beryllium powder and alloy samples were also declared to range from about 8 to 28 wt % uranium. The chemical and uranium isotopic analyses done at the Y-12 Plant confirm the Ulba plant declarations. All samples appear to have been enriched using some reprocessed uranium, probably from recovery of uranium from plutonium production reactors. As a result, all samples contain some {sup 236}U and {sup 232}U and have small but measurable quantities of plutonium. This plutonium could be the result of either contamination carried over from the enrichment process or cross-contamination from weapons material. It is not the result of direct reactor exposure. Neither the {sup 232}U nor the plutonium concentrations are sufficiently high to provide a significant industrial health hazard. Both are well within established or proposed acceptance criteria for storage at Y-12. The trace metal analyses showed that, with the exception of beryllium, there are no trace metals in any of these HEU samples that pose a significant health hazard.

Gift, E.H.

1995-05-01T23:59:59.000Z

119

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

SciTech Connect

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

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

2006-07-01T23:59:59.000Z

120

Comparison of the FRM-II HEU design with an alternative LEU design  

SciTech Connect

The FRM-II reactor design of the Technical University of Munich has a compact core that utilizes fuel plates containing highly-enriched uranium (HEU, 93%). This paper presents an alternative core design utilizing low-enriched uranium (LEU, <20%) silicide fuel with 4.8 g/cm{sup 3} that provides nearly the same neutron flux for experiments as the HEU design, but has a less favorable fuel cycle economy. If an LEU fuel with a uranium density of 6.0 - 6.5 g/cm{sup 3} were developed, the alternative design would provide the same neutron flux and use the same number of cores per year as the HEU design. The results of this study show that there are attractive possibilities for using LEU fuel instead of HEU fuel in the FRM-II. Further optimization of the LEU design and near-term availability of LEU fuel with a uranium density greater than 4.8 g/cm{sup 3} would enhance the performance of the LEU core. The RERTR Program is ready to exchange information with the Technical University of Munich to resolve any differences that may exist and to identify design modifications that would optimize reactor performance utilizing LEU fuel.

Mo, S.C.; Hanan, N.A.; Matos, J.E.

1995-12-01T23:59:59.000Z

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121

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

SciTech Connect

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

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

2003-06-13T23:59:59.000Z

122

Nuclear Materials Identification System (NMIS) with Gamma Spectrometry for Attributes of Pu, HEU, and Detection of HE and Chemical Agents  

SciTech Connect

A combined Nuclear Materials Identification System (NMIS)-gamma ray spectrometry system can be used passively to obtain the following attributes of Pu: presence, fissile mass, 240/239 ratio, and metal vs. oxide. This system can also be used with a small, portable, DT neutron generator to measure the attributes of highly enriched uranium (HEU): presence, fissile mass, enrichment, metal vs. oxide; and detect the presence of high explosives (HE). For the passive system, time-dependent coincidence distributions can be used for the presence, fissile mass, metal vs. oxide for Pu, and gamma-ray spectrometry can be used for 239/240 ratio and presence. So presence can be confirmed by two methods. For the active system with a DT neutron generator, all four attributes for both Pu and HEU can be determined from various features of the time-dependent coincidence distribution measurements for both Pu and HEU. Active gamma ray spectrometry would also give presence and 240/239 ratio for Pu, enrichment for HEU, and metal vs. oxide for both. Active gamma ray spectrometry would determine the presence of HE. The various features of time-dependent coincidence distributions and gamma ray spectrometry that determine these attributes are discussed with some examples from previous determinations.

Mihalczo, J. T.; Mattingly, J. K.; Mullens, J. A.; Neal, J. S.

2002-05-01T23:59:59.000Z

123

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

Science Conference Proceedings (OSTI)

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

NONE

1995-05-01T23:59:59.000Z

124

Ion-induced gammas for photofission interrogation of HEU.  

SciTech Connect

High-energy photons and neutrons can be used to actively interrogate for heavily shielded special nuclear material (SNM), such as HEU (highly enriched uranium), by detecting prompt and/or delayed induced fission signatures. In this work, we explore the underlying physics for a new type of photon source that generates high fluxes of mono-energetic gamma-rays from low-energy (<500 keV) proton-induced nuclear reactions. The characteristic energies (4- to 18-MeV) of the gamma-rays coincide with the peak of the photonuclear cross section. The source could be designed to produce gamma-rays of certain selected energies, thereby improving the probability of detecting shielded HEU or providing a capability to determine enrichment inside sealed containers. The fundamental physics of such an interrogation source were studied in this LDRD through scaled ion accelerator experiments and radiation transport modeling. The data were used to assess gamma and neutron yields, background, and photofission-induced signal levels from several (p,{gamma}) target materials under consideration.

Doyle, Barney Lee (Sandia National Laboratories, Albuquerque, NM); Antolak, Arlyn J.; Morse, Daniel H.; Provencio, Paula Polyak (Sandia National Laboratories, Albuquerque, NM)

2006-03-01T23:59:59.000Z

125

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

SciTech Connect

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

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

2006-07-01T23:59:59.000Z

126

Surplus Highly Enriched Uranium Disposition Program plan  

SciTech Connect

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

1996-10-01T23:59:59.000Z

127

Planning the HEU to LEU Transition for the NBSR  

SciTech Connect

A study has been carried out to understand how the NIST research reactor (NBSR) might be converted from using high-enriched uranium (HEU) to using low-enriched uranium (LEU) fuel. An LEU fuel design had previously been determined which provides an equilibrium core with the desirable fuel cycle length - a very important parameter for maintaining the experimental, scientific program supported by the NBSR. In the present study two options for getting to the equilibrium state are considered. One option starts with the loading of an entire core of fresh fuel. This was determined to be unacceptable. The other option makes use of the current fuel management scheme wherein four fresh fuel elements are loaded at the beginning of each cycle. However, it is shown that without some alterations to the fuel cycle, none of the transition cores containing both HEU and LEU fuel have sufficient excess reactivity to operate the reactor for the optimum length. It was determined that operating the first mixed cycle for a sufficiently reduced length of time provides the excess reactivity which enables subsequent cycles to be run for the desired number of days.

Hanson, A.L.; Diamond, D.

2011-09-12T23:59:59.000Z

128

Planning the HEU to LEU Transition for the NBSR  

SciTech Connect

A study has been carried out to understand how the NIST research reactor (NBSR) might be converted from using high-enriched uranium (HEU) to using low-enriched uranium (LEU) fuel. An LEU fuel design had previously been determined which provides an equilibrium core with the desirable fuel cycle length—a very important parameter for maintaining the experimental, scientific program supported by the NBSR. In the present study two options for getting to the equilibrium state are considered. One option starts with the loading of an entire core of fresh fuel. This was determined to be unacceptable. The other option makes use of the current fuel management scheme wherein four fresh fuel elements are loaded at the beginning of each cycle. However, it is shown that without some alterations to the fuel cycle, none of the transition cores containing both HEU and LEU fuel have sufficient excess reactivity to enable reactor operation for the required amount of time. It was determined that operating the first mixed cycle for a sufficiently reduced length of time provides the excess reactivity which enables subsequent transition cycles to be run for the desired number of days.

Hanson, A.L.; Diamond, D.

2011-10-24T23:59:59.000Z

129

HEU to LEU conversion and blending facility: Oxide blending alternative to produce LEU oxide for commercial use  

SciTech Connect

The United States Department of Energy (DOE) is examining options for the disposition of surplus weapons-usable fissile materials and storage of all weapons-usable fissile materials. Disposition is a process of use or disposal of material that results in the material being converted to a form that is substantially and inherently more proliferation-resistant than the original form. Examining options for increasing the proliferation resistance of highly enriched uranium (HEU) is part of this effort. This document provides data to be used in the environmental impact analysis for the oxide blending HEU disposition option. This option provides for a yearly HEU throughput of 1 0 metric tons (MT) of uranium metal with an average U235 assay of 50% blended with 165 MT of natural assay triuranium octoxide (U{sub 3} O{sub 8}) per year to produce 177 MT of 4% U235 assay U{sub 3} O{sub 8}, for LWR fuel. Since HEU exists in a variety of forms and not necessarily in the form to be blended, worst case scenarios for preprocessing prior to blending will be assumed for HEU feed streams.

1995-09-01T23:59:59.000Z

130

HPGe Detector Efficiency Calibration Using HEU Standards  

SciTech Connect

The Analytical Development Section of 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 Solid Waste's Waste Acceptance Criteria, Material Control and Accountability, and to meet criticality safety controls. Two measurement systems will be used to determine highly enriched uranium (HEU) holdup: One is a portable HPGe detector and EG and G Dart system that contains high voltage power supply and signal processing electronics. A personal computer with Gamma-Vision software was used to provide an MCA card, and space to store and manipulate multiple 4096-channel g-ray spectra. The other is a 2 inches x 2 inches NaI crystal with an MCA that uses a portable computer with a Canberra NaI plus card installed. This card converts the PC to a full function MCA and contains the ancillary electronics, high voltage power supply and amplifier, required for data acquisition. This report describes and documents the HPGe point, line, area, and constant geometry-constant transmission detector efficiency calibrations acquired and calculated for use in conducting holdup measurements as part of the overall deactivation project of building 321-M.

Salaymeh, S.R.

2000-10-12T23:59:59.000Z

131

Prompt Neutron Time Decay in Single HEU and DU Metal Annular Storage Castings  

SciTech Connect

Previous measurements of highly enriched uranium (HEU) storage castings performed by Oak Ridge National Laboratory (ORNL) at the Y-12 National Security Complex showed a prompt neutron time decay that is not exponential. These measurements showed that multiple time constants originating from multiplication, time-of-flight, scattering in the assembly and room return could be associated with this prompt neutron decay. In this work, the contribution not associated with neutron multiplication was investigated via measurements with a depleted uranium (DU) casting. The measurements at ORNL used an annular (5.0-in OD, 3.5-in ID, 6.0-in H) DU casting with a time-tagged 252Cf source, centered vertically on the axis, and four closely coupled 1 1 6-in.-long plastic scintillators with -in.- thick lead shielding adjacent to the outer surface of the casting. This setup was identical to the configuration used in the previously performed measurements with HEU castings at Y-12. The time correlation between fission events and detections in the plastic scintillators was measured, as well as the time distribution of coincidences between multiple detectors within a 512-ns time window. The measurement results were then compared to MCNP-PoliMi calculations and the previous HEU measurements. Time constants from decay fits to the HEU and DU data were compared to characterize the contributions resulting from multiplication, time-of-flight, and scattering.

Pena, Kirsten E [ORNL; McConchie, Seth M [ORNL; Mihalczo, John T [ORNL

2010-01-01T23:59:59.000Z

132

Passive Time Coincidence Measurements with Assemblies of HEU and DU Castings  

SciTech Connect

Five highly enriched (93.2 wt% 235U) uranium (HEU) and three depleted (0.2 wt% 235U) uranium (DU) castings were available for passive time coincidence measurements with two arrays of eight total liquid scintillation detectors and two arrays of 32 total moderated 3He detectors. In addition to measurements with each type of casting, measurements were also performed with both types of castings intermingled. The spontaneous fission rate for the casting is primarily due to the presence of 238U. Thus, the DU castings have ~15 times more spontaneous fission that the HEU castings with very little addition in multiplication with each addition of a DU casting. The characteristic time behavior of the fission chain multiplication process is determined by the 235U. The presence of the DU castings in these assemblies has three physical effects: provision of an additional source of neutrons to initiate fission chains in the HEU, reflection of neutrons back into the HEU, and shielding of the detectors from fission chain neutrons and gammas. The effects are investigated within the context of the time coincidence distributions using the signal-triggered multiplets and Feynman variances.

McConchie, Seth M [ORNL; Hausladen, Paul [ORNL; Mihalczo, John T [ORNL

2009-01-01T23:59:59.000Z

133

Analyses for conversion of the Georgia Tech Research Reactor from HEU to LEU fuel  

SciTech Connect

The 5 MW Georgia Tech Research Reactor (GTRR) is a heterogeneous, heavy water moderated and cooled reactor, fueled with highly-enriched uranium aluminum alloy fuel plates. The GTRR is required to convert to low enrichment (LEU) fuel in accordance with USNRC policy. Results of design and safety analyses performed by the RERTR Program at the Argonne National Laboratory for LEU conversion of the GTRR are summarized. Only those parameters which could change as a result of replacing the fuel are addressed. The performance of the reactor and all safety margins with LEU fuel are expected to be about the same as those with the current HEU fuel.

Matos, J.E.; Mo, S.C.; Woodruff, W.L.

1992-01-01T23:59:59.000Z

134

Analyses for conversion of the Georgia Tech Research Reactor from HEU to LEU fuel  

Science Conference Proceedings (OSTI)

The 5 MW Georgia Tech Research Reactor (GTRR) is a heterogeneous, heavy water moderated and cooled reactor, fueled with highly-enriched uranium aluminum alloy fuel plates. The GTRR is required to convert to low enrichment (LEU) fuel in accordance with USNRC policy. Results of design and safety analyses performed by the RERTR Program at the Argonne National Laboratory for LEU conversion of the GTRR are summarized. Only those parameters which could change as a result of replacing the fuel are addressed. The performance of the reactor and all safety margins with LEU fuel are expected to be about the same as those with the current HEU fuel.

Matos, J.E.; Mo, S.C.; Woodruff, W.L.

1992-12-31T23:59:59.000Z

135

Disposition of Unirradiated Sodium Bonded EBR-II Driver Fuel Elements and HEU Scrap: Work Performed for FY 2007  

SciTech Connect

Specific surplus high enriched uranium (HEU) materials at the Idaho National Laboratory (INL) Materials and Fuels Complex (MFC) will be transferred to a designated off-site receiving facility. The DOE High Enriched Uranium Disposition Program Office (HDPO) will determine which materials, if any, will be prepared and transferred to an off-site facility for processing and eventual fabrication of fuel for nuclear reactors. These surplus HEU materials include approximately 7200 kg unirradiated sodium-bonded EBR-II driver fuel elements, and nearly 800 kg of HEU casting scrap from the process which formed various sodium-bonded fuels (including the EBR-II driver elements). Before the driver fuel can be packaged for shipment, the fuel elements will require removal of the sodium bond. The HEU scrap will also require repackaging in preparation for off-site transport. Preliminary work on this task was authorized by BWXT Y-12 on Nov 6, 2006 and performed in three areas: • Facility Modifications • Safety Documentation • Project Management

Karen A Moore

2007-04-01T23:59:59.000Z

136

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

SciTech Connect

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

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

1980-12-01T23:59:59.000Z

137

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

SciTech Connect

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

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

2009-04-01T23:59:59.000Z

138

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

Science Conference Proceedings (OSTI)

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

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

2012-07-01T23:59:59.000Z

139

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

SciTech Connect

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

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

2002-05-23T23:59:59.000Z

140

Disposition of Surplus Highly Enriched Uranium  

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

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

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


141

Transition from HEU to LEU fuel in Romania`s 14-MW TRIGA reactor  

SciTech Connect

The 14-MW TRIGA steady state reactor (SSR) located in Pitesti, Romania, first went critical in the fall of 1979. Initially, the core configuration for full power operation used 29 fuel clusters each containing a 5 {times} 5 square array of HEU (10 wt%) -- ZrH -- Er (2.8 wt%) fuel-moderator rods (1.295 cm o.d.) clad in Incology. With a total inventory of 35 HEU fuel clusters, burnup considerations required a gradual expansion of the core from 29 to 32 and finally to 35 clusters before the reactor was shut down because of insufficient excess reactivity. At this time each of the original 29 fuel clusters had an overage {sup 235}U burnup in the range from 50 to 62%. Because of the US policy regarding the export of highly enriched uranium, fresh HEU TRIGA replacement fuel is not available. After a number of safety-related measurements, the SSR is expected to resume full power operation in the near future using a mixed core containing five LEU TRIGA clusters of the same geometry as the original fuel but with fuel-moderator rods containing 45 wt% U (19.7% {sup 235}U enrichment) and 1.1 wt% Er. Rods for 14 additional LEU fuel clusters will be fabricated by General Atomics. In support of the SSR mixed core operation numerous neutronic calculations have been performed. This paper presents some of the results of those calculations.

Bretscher, M.M.; Snelgrove, J.L.

1991-12-31T23:59:59.000Z

142

Coordination Between the HEU Transparency Program and the Material Protection, Control and Accountability Program  

SciTech Connect

DOE sponsored programs such as Material Protection Control and Accountability (MPC&A) and implementation of the Highly-Enriched Uranium (HEU) Transparency Program send US personnel into Russian nuclear facilities and receive Russian representatives from these programs. While there is overlap in the Russian nuclear facilities visited by these two programs, there had not been any formal mechanism to share information between them. Recently, an MPC&A/HEU Working Group was developed to facilitate the sharing of appropriate information and to address concerns expressed by Minatom and Russian facility personnel such as US visit scheduling conflicts. This paper discusses the goals of the Working Group and ways it has helped to allow the programs to work more efficiently with the Russian facilities.

Glaser, J.; Hernandez, J.; Dougherty, D.; Bieniawski, A.; Cahalane, P.; Mastal, E.

2000-06-30T23:59:59.000Z

143

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

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

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

144

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

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

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

145

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

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

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

146

Low enrichment fuel conversion for Iowa State University. Final report  

SciTech Connect

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

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

1996-10-17T23:59:59.000Z

147

HEU to LEU Conversion and Blending Facility: UF{sub 6} blending alternative to produce LEU UF{sub 6} for commercial use  

Science Conference Proceedings (OSTI)

US DOE is examining options for disposing of surplus weapons-usable fissile materials and storage of all weapons-usable fissile materials; the nuclear material will be converted to a form more proliferation- resistant than the original form. Examining options for increasing the proliferation resistance of highly enriched uranium (HEU) is part of this effort. Five technologies for blending HEU will be assessed; blending as UF{sub 6} to produce a UF{sub 6} product for commercial use is one of them. This document provides data to be used in the environmental impact analysis for the UF{sub 6} blending HEU disposition option. Resource needs, employment needs, waste and emissions from plant, hazards, accident scenarios, and intersite transportation are discussed.

NONE

1995-09-01T23:59:59.000Z

148

Plutonium isotopic analysis of highly enriched mixed oxides  

SciTech Connect

We investigated the analysis method used by the International Atomic Energy Agency (IAEA) to determine the plutonium isotopic composition of highly enriched mixed oxides (MOX). The IAEA currently uses the Cicero multichannel analyzer and the IAEAPU algorithm for its analysis. In our investigation the plutonium isotopic measurements were found to be good for PuO/sub 2/ powder or low-enriched MOX, but acceptable for highly enriched MOX in IAEA special nuclear material (SNM) accountability applications. The gamma-ray interferences from /sup 235/U resulted in underestimation of the isotopic composition of /sup 239/Pu and overestimation of all other plutonium isotopes. Samples with high /sup 240/Pu content were found to have significantly higher error in plutonium isotopic analyses of highly enriched MOX. Code modifications or use of calibration curves are necessary for plutonium isotopic analyses of highly enriched MOX in IAEA SNM accountability applications.

Clement, S.D.; Augustson, R.H.

1986-08-01T23:59:59.000Z

149

Fissile Mass Flow Monitor Implementation for Transparency in HEU Blenddown at the URAL Electrochemical Integrated Plant (UEIP) in Novouralsk  

SciTech Connect

The Oak Ridge National Laboratory (ORNL) Fissile Mass Flow Monitor (FMFM) was deployed at the Ural Electrochemical Integrated Plant (UEIP) highly enriched uranium (HEU) blending facility in January and February 1999 at Novouralsk in Russia for the DOE HEU Transparency Program. The FMFM provides unattended monitoring of the fissile mass flow of the uranium hexafluoride (UF{sub 6}) gas in the process lines of HEU, the low enriched uranium (LEU) blend stock, and the product LEU (P-LEU) of the blending tee non-intrusively. To do this, uranium-235 (U-235) fissions are induced in the UF{sub 6} by a thermalized and modulated californium-252 (Cf-252) neutron source placed on each process line. A set of detectors, located downstream of source, measure delayed gamma rays emitted by the resulting fission fragments. The observed delay in the time correlated measurement between the source and the detector signal provides the velocity of UF{sub 6} and its amplitude is related to the U- 235 content in UF{sub 6}. An on-line computer controls the source modulator, processes the collected detector data, and displays the results. The UEIP Main and the Reserved process lines were implemented with minor modifications. The FMFM monitors the HEU blending operation by measuring UF{sub 6} flows in the process blending lines, and the traceability of the HEU flow from the blend point to the P-LEU. The detail operational characteristics of the FMFM software (FM2) and the measurement methodology used are presented.

March-Leuba, J.; Mastal, E.; Powell, D.; Sumner, J.; Uckan, T.; Vines, B.

1999-07-25T23:59:59.000Z

150

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

National Nuclear Security Administration (NNSA)

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

151

Feasibility study for early removal of HEU from CPP-651-Phase II  

SciTech Connect

A two-phase feasibility study was initiated in late 1996 to identify a way to expedite the removal of SNM from the CPP-651 vault. The first phase of this study provided preliminary information that appeared promising, but needed additional detailed planning and evaluate to validate the concepts and conclusions. The focus of Phase 2 was to provide the validation via resource-loaded schedules and more detailed cost estimates. Section 1 describes the purpose and objectives of the Phase 2 tasks and the programmatic drivers that influence related CPP-651 high-enriched uranium (HEU) management issues. Section 2 identifies the evaluation criteria and methodology and the transfer issues and barriers preventing shipment. Section 3 provides site-specific background information for the CPP-651 facility and the Idaho National Engineering and Environmental Laboratory (INEEL) and describes the development of the basic material removal schedule, the proposed base case plan for removal of SNM, and the proposed HEU material management/shipping issues and strategies. Section 4 identifies the proposed options for accelerated removal of SNM and how they were evaluated via detailed scheduling, resource histograms, and cost analysis. Section 5 summarizes principal tasks for implementing this plan and other related HEU CPP-651 management issues that require continued planning efforts to assure successful implementation of this proposed early removal strategy.

Smith, C.V.; Henry, R.; Milligan, C.; Harmon, B.; Peterson, J.; Thom, M.A.; Campbell, R.; Hendrix, B.

1997-09-01T23:59:59.000Z

152

HEU Holdup Measurements on 321-M A-Lathe  

Science Conference Proceedings (OSTI)

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 enriched 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.

Dewberry, R.A.

2002-04-30T23:59:59.000Z

153

The Rhode Island Nuclear Science Center conversion from HEU to LEU fuel  

SciTech Connect

The 2-MW Rhode Island Nuclear Science Center (RINSC) open pool reactor was converted from 93% UAL-High Enriched Uranium (HEU) fuel to 20% enrichment U3Si2-AL Low Enriched Uranium (LEU) fuel. The conversion included redesign of the core to a more compact size and the addition of beryllium reflectors and a beryllium flux trap. A significant increase in thermal flux level was achieved due to greater neutron leakage in the new compact core configuration. Following the conversion, a second cooling loop and an emergency core cooling system were installed to permit operation at 5 MW. After re-licensing at 2 MW, a power upgrade request will be submitted to the NRC.

Tehan, Terry

2000-09-27T23:59:59.000Z

154

2 x 2 Polyethylene Reflected and Moderated Highly Enriched Uranium System with Rhenium  

SciTech Connect

The 2 × 2 array HEU-Re experiment was performed on the Planet universal critical assembly machine on November 4th, 2003 at the Los Alamos Critical Experiments Facility (LACEF) at Los Alamos National Laboratory (LANL). For this experiment, there were 10 ˝ units, each full unit containing four HEU foils and two rhenium foils. The top unit contained only two HEU foils and two rhenium foils. A total of 42 HEU foils were used for this experiment. Rhenium is a desirable cladding material for space nuclear power applications. This experiment consisted of HEU foils interleaved with rhenium foils and is moderated and reflected by polyethylene plates. A unit consisted of a polyethylene plate, which has a recess for rhenium foils, and four HEU foils in a single layer in the top recess of each polyethylene plate. The Planet universal criticality assembly machine has been previously used in experiments containing HEU foils interspersed with SiO2 (HEU-MET-THERM-001), Al (HEU-MET-THERM-008), MgO (HEU-MET-THERM-009), Gd foils (HEU-MET-THERM-010), 2 × 2 × 26 Al (HEU-MET-THERM-012), Fe (HEU-MET-THERM-013 and HEU-MET-THERM-015), 2 × 2 × 23 SiO2 (HEU-MET-THERM-014), 2 × 2 × 11 hastalloy plates (HEU-MET-THERM-016), and concrete (HEU-MET-THERM-018). The 2 × 2 array of HEU-Re is considered acceptable for use as a benchmark critical experiment.

A. Nichole Ellis; Jesson Hutchinson; John D. Bess; Dmitry N. Polyakov; Evgeny S. Glushkov; Alexey E. Glushkov

2010-09-01T23:59:59.000Z

155

Analysis of Accidents at the Pakistan Research Reactor-1 Using Proposed Mixed-Fuel (HEU and LEU) Core  

Science Conference Proceedings (OSTI)

The Pakistan Research Reactor-1 (PARR-1) was converted from highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel in 1991. The reactor is running successfully, with an upgraded power level of 10 MW. To save money on the purchase of costly fresh LEU fuel elements, the use of less burnt HEU spent fuel elements along with the present LEU fuel elements is being considered. The proposal calls for the HEU fuel elements to be placed near the thermal column to gain the required excess reactivity. In the present study the safety analysis of a proposed mixed-fuel core has been carried out at a calculated steady-state power level of 9.8 MW. Standard computer codes and correlations were employed to compute various parameters. Initiating events in reactivity-induced accidents involve various modes of reactivity insertion, namely, start-up accident, accidental drop of a fuel element on the core, flooding of a beam tube with water, and removal of an in-pile experiment during reactor operation. For each of these transients, time histories of reactor power, energy released, temperature, and reactivity were determined.

Bokhari, Ishtiaq H. [Pakistan Institute of Nuclear Science and Technology (Pakistan)

2004-12-15T23:59:59.000Z

156

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

SciTech Connect

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

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

2007-10-01T23:59:59.000Z

157

Profile of World Uranium Enrichment Programs-2009  

Science Conference Proceedings (OSTI)

It is generally agreed that the most difficult step in building a nuclear weapon is acquiring fissile material, either plutonium or highly enriched uranium (HEU). Plutonium is produced in a nuclear reactor, whereas HEU is produced using a uranium enrichment process. Enrichment is also an important step in the civil nuclear fuel cycle, in producing low enriched uranium (LEU) for use as fuel for nuclear reactors to generate electricity. However, the same equipment used to produce LEU for nuclear reactor fuel can also be used to produce HEU for weapons. Safeguards at an enrichment plant are the array of assurances and verification techniques that ensure uranium is not diverted or enriched to HEU. There are several techniques for enriching uranium. The two most prevalent are gaseous diffusion, which uses older technology and requires a lot of energy, and gas centrifuge separation, which uses more advanced technology and is more energy efficient. Gaseous diffusion plants (GDPs) provide about 40% of current world enrichment capacity but are being phased out as newer gas centrifuge enrichment plants (GCEPs) are constructed. Estimates of current and future enrichment capacity are always approximate, due to the constant upgrades, expansions, and shutdowns occurring at enrichment plants, largely determined by economic interests. Currently, the world enrichment capacity is approximately 56 million kilogram separative work units (SWU) per year, with 22.5 million in gaseous diffusion and more than 33 million in gas centrifuge plants. Another 34 million SWU/year of capacity is under construction or planned for the near future, almost entirely using gas centrifuge separation. Other less-efficient techniques have also been used in the past, including electromagnetic and aerodynamic separations, but these are considered obsolete, at least from a commercial perspective. Laser isotope separation shows promise as a possible enrichment technique of the future but has yet to be demonstrated commercially. In the early 1980s, six countries developing gas centrifuge technology (United States, United Kingdom, Germany, the Netherlands, Japan, and Australia) along with the International Atomic Energy Agency and the European Atomic Energy Community began developing effective safeguards techniques for GCEPs. This effort was known as the Hexapartite Safeguards Project (HSP). The HSP had the goal of maximizing safeguards effectiveness while minimizing the cost to the operator and inspectorate, and adopted several recommendations, such as the acceptance of limited-frequency unannounced access inspections in cascade halls, and the use of nondestructive assay measurements and tamper-indicating seals. While only the HSP participants initially committed to implementing all the measures of the approach, it has been used as a model for the safeguards applied to GCEPs in additional states. Uranium enrichment capacity has continued to expand on all fronts in the last few years. GCEP capacity is expanding in anticipation of the eventual shutdown of the less-efficient GDPs, the termination of the U.S.-Russia HEU blend-down program slated for 2013, and the possible resurgence of nuclear reactor construction as part of an expected 'Nuclear Renaissance'. Overall, a clear trend in the world profile of uranium enrichment plant operation is the continued movement towards multinational projects driven by commercial and economic interests. Along this vein, the safeguards community is continuing to develop new safeguards techniques and technologies that are not overly burdensome to enrichment plant operators while delivering more effective and efficient results. This report provides a snapshot overview of world enrichment capacity in 2009, including profiles of the uranium enrichment programs of individual states. It is a revision of a 2007 report on the same topic; significant changes in world enrichment programs between the previous and current reports are emphasized. It is based entirely on open-source information, which is dependent on published sources and may theref

Laughter, Mark D [ORNL

2009-04-01T23:59:59.000Z

158

Transient analyses for the Uzbekistan VVR-SM reactor with IRT-3M HEU fuel and IRT-4M LEU fuel : ANL independent verification results.  

Science Conference Proceedings (OSTI)

Calculations have been performed for postulated transients in the VVR-SM Reactor at the Institute of Nuclear Physics (INP) of the Academy of Sciences in the Republic of Uzbekistan. (The reactor designation in Cyrillic is BBP-CM; transliterating characters to English gives VVRSM but translating words gives WWR-SM.) These calculations have been performed at the request of staff of the INP who are performing similar calculations. The transients considered were established during working meetings between Argonne National Laboratory (ANL) and INP staff during summer 2006 [Ref. 1], subsequent email correspondence, and subsequent staff visits. Calculations were performed for the current high-enriched uranium (HEU) core, the proposed low-enriched uranium (LEU) core, and one mixed HEU-LEU core during the transition. These calculations have been performed independently from those being performed by INP and serve as one step in the verification process.

Garner, P. L.; Hanan, N. A.; Nuclear Engineering Division

2007-09-24T23:59:59.000Z

159

Efficiency Calibration Using HEU Standards of 2-Inch by 2-Inch NaI Detector  

SciTech Connect

The Analytical Development Section of SRTC was requested by the Facilities Disposition Division (FDD) to determine the holdup of highly enriched uranium (HEU) 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 solid waste Waste Acceptance Criteria, Material Control and Accountability, and to meet criticality safety controls. Two measurement systems will be used to determine HEU holdup: One is a portable EG and G Dart system that contains Gamma-Vision software to support a Multichannel Analyzer (MCA) card, high voltage power, and space to store and manipulate multiple 4096-channel gamma-ray spect ra. The other is a 2-inch x 2-inch NaI crystal with an MCA that uses a portable computer with a Canberra NaI plus card installed. This card converts the PC to a full function MCA and contains the ancillary electronics, high voltage power supply and amplifier, required for data acquisition. This report will discuss the calibration of the 2-inch x 2-inch NaI detector.

Dewberry, R. A.

2000-10-24T23:59:59.000Z

160

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

Science Conference Proceedings (OSTI)

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

Mihalczo, John T [ORNL

2011-01-01T23:59:59.000Z

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


161

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

National Nuclear Security Administration (NNSA)

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

162

Alternative dispositioning methods for HEU spent nuclear fuel at the Savannah River Site  

SciTech Connect

The United States has a strong policy on prevention of the international spread of nuclear weapons. This policy was announced in Presidential Directive PDD-13 and summarized in a White House press release September 27, 1993. Two cornerstones of this policy are: seek to eliminate where possible the accumulation of stockpiles of highly- enriched uranium or plutonium; propose{hor_ellipsis}prohibiting the production of highly-enriched uranium (HEU) or plutonium for nuclear explosives purposes or outside international safeguards. The Department of Energy is currently struggling to devise techniques that safely and efficiently dispose of spent nuclear fuel (SNF) while satisfying national non-proliferation policies. SRS plans and proposals for disposing of their SNF are safe and cost effective, and fully satisfy non-proliferation objectives.

Krupa, J.F.; McKibben, J.M.; Parks, P.B.; DuPont, M.E.

1995-11-01T23:59:59.000Z

163

Power distributions in fresh and depleted LEU and HEU cores of the MITR reactor.  

Science Conference Proceedings (OSTI)

The Massachusetts Institute of Technology Reactor (MITR-II) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. It delivers a neutron flux comparable to current LWR power reactors in a compact 6 MW core using Highly Enriched Uranium (HEU) fuel. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context, most research and test reactors both domestic and international have started a program of conversion to the use of Low Enriched Uranium (LEU) fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (UMo) is expected to allow the conversion of U.S. domestic high performance reactors like the MITR-II reactor. Toward this goal, core geometry and power distributions are presented. Distributions of power are calculated for LEU cores depleted with MCODE using an MCNP5 Monte Carlo model. The MCNP5 HEU and LEU MITR models were previously compared to experimental benchmark data for the MITR-II. This same model was used with a finer spatial depletion in order to generate power distributions for the LEU cores. The objective of this work is to generate and characterize a series of fresh and depleted core peak power distributions, and provide a thermal hydraulic evaluation of the geometry which should be considered for subsequent thermal hydraulic safety analyses.

Wilson, E.H.; Horelik, N.E.; Dunn, F.E.; Newton, T.H., Jr.; Hu, L.; Stevens, J.G. (Nuclear Engineering Division); (2MIT Nuclear Reactor Laboratory and Nuclear Science and Engineering Department)

2012-04-04T23:59:59.000Z

164

Neutronic Analyses for HEU to LEU fuel conversion of the Massachusetts Institute of Technology.  

SciTech Connect

The Massachusetts Institute of Technology (MIT) reactor (MITR-II), based in Cambridge, Massachusetts, is a research reactor designed primarily for experiments using neutron beam and in-core irradiation facilities. It delivers a neutron flux comparable to current LWR power reactors in a compact 6 MW core using Highly Enriched Uranium (HEU) fuel. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context, most research and test reactors both domestic and international have started a program of conversion to the use of Low Enriched Uranium (LEU) fuel. A new type of LEU fuel based on a mixture of uranium and molybdenum (UMo) is expected to allow the conversion of compact high performance reactors like the MITR-II. This report presents the results of steady state neutronic safety analyses for conversion of MITR-II from the use of HEU fuel to the use of U-Mo LEU fuel. The objective of this work was to demonstrate that the safety analyses meet current requirements for an LEU core replacement of MITR-II.

Wilson, E. H.; Newton, T. H.; Bergeron, A.; Horelik, N.; Stevens, J. G (Nuclear Engineering Division); ( NS)

2011-03-02T23:59:59.000Z

165

Sealing of process valves for the HEU downblending verification experiment at Portsmouth  

SciTech Connect

At the Portsmouth Gaseous Diffusion Plant in Piketon, Ohio, USA, excess inventory of highly-enriched uranium (HEU) from US defense programs is being diluted to low-enriched uranium (LEU) for commercial use. The conversion is subject to a Verification Experiment overseen by the International Atomic Energy Agency (IAEA). The Verification Experiment is making use of monitoring technologies developed and installed by several DOE laboratories. One of the measures is a system for sealing valves in the process piping, which secures the path followed by uranium hexafluoride gas (UF{sub 6}) from cylinders at the feed stations to the blend point, where the HEU is diluted with LEU. The Authenticated Item Monitoring System (AIMS) was the alternative proposed by Sandia National Laboratories that was selected by the IAEA. Approximately 30 valves were sealed by the IAEA using AIMS fiber-optic seals (AFOS). The seals employ single-core plastic fiber rated to 125 C to withstand the high-temperature conditions of the heated piping enclosures at Portsmouth. Each AFOS broadcasts authenticated seal status and state-of-health messages via a tamper-protected radio-frequency transmitter mounted outside of the heated enclosure. The messages are received by two collection stations, operated redundantly.

Baldwin, G.T.; Bartberger, J.C.; Jenkins, C.D.; Perlinski, A.W.; Schoeneman, J.L. [Sandia National Labs., Albuquerque, NM (United States); Gordon, D.M. [Brookhaven National Lab., Upton, NY (United States); Whiting, N.E. [International Atomic Energy Agency (IAEA); Bonner, T.N.; Castle, J.M. [Lockheed Martin Utility Services, Piketon, OH (United States). Portsmouth Gaseous Diffusion Plant

1998-07-01T23:59:59.000Z

166

Basic characterization of highly enriched uranium by gamma spectrometry  

E-Print Network (OSTI)

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

Nguyen, C T

2006-01-01T23:59:59.000Z

167

Basic characterization of highly enriched uranium by gamma spectrometry  

E-Print Network (OSTI)

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

Cong Tam Nguyen; Jozsef Zsigrai

2005-08-25T23:59:59.000Z

168

Analyses for conversion of the Georgia Tech Research Reactor from HEU to LEU fuel  

SciTech Connect

This document presents information concerning: analyses for conversion of the Georgia Tech Research Reactor from HEU to LEU; changes to technical specifications mandated by the conversion of the GTRR to low enrichment fuel; changes in the Safety Analysis Report mandated by the conversion of the GTRR to low enrichment fuel; and copies of all changed pages of the SAR and the technical specifications.

Matos, J.E.; Mo, S.C.; Woodruff, W.L.

1992-09-01T23:59:59.000Z

169

Nickel container of highly-enriched uranium bodies and sodium  

SciTech Connect

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

Zinn, Walter H. (Hinsdale, IL)

1976-01-01T23:59:59.000Z

170

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

SciTech Connect

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

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

2002-05-17T23:59:59.000Z

171

Subcriticality Measurements with HEU (93.2) Metal Annular Storage Castings  

SciTech Connect

These carefully performed and documented measurements with unreflected and unmoderated highly enriched uranium (HEU) castings can be used to benchmark calculational methods for the time decay of the fission chain multiplication process as measured with small (1 x 1 x 6 in. thick plastic scintillators with 1/4-in.-thick lead on all detector surfaces) detectors adjacent to the tightly fitting stainless steel cans that contained the HEU ({approx}93 wt%) metal. Prompt time decay measurements were performed stimulating the fission chain multiplication process with a timed, tagged Cf spontaneous fission source that emitted fission-spectrum neutrons and a time and directionally tagged 14.1-MeV neutrons from the DT reaction in a steady state generator with an embedded alpha detector. Time decay measurements were performed with HEU masses varying from 18 to 90 kg for a wide variety of source-detector-casting configurations. The use of a DT generator provided no addition information about the fission chain behavior beyond that provided by a time-tagged Cf spontaneous fission source. The main quantities obtained in the measurements were (1) the time distribution of the counts in a detector after a neutron fission in the Cf source or after the alpha detection coincident with the emission of a neutron from the DT generator (the equivalent of a pulsed neutron measurement with a randomly pulsed source) and (2) the time distribution of counts in one detector after a count in another detector (the equivalent of a two-detector Rossi-alpha measurement). Monte Carlo calculations using the MCNP-PoliMi coupled gamma-neutron transport code generally agreed with the measurement results except for some differences early in the fission chain decay process. The measurements that were performed with the HEU about 1 m above the floor were considerably affected by room return neutrons at times as early as 100 ns, and at times after 300 ns, a major portion of the time response was associated with the interaction of the HEU assemblies with the floor. This room-return effect increased with the size of the assembly because the larger assemblies subtend a larger solid angle to a neutron returning from the floor.

Henkel, James J [ORNL; Wright, Michael C [ORNL; Archer, Daniel E [ORNL; Mullens, James Allen [ORNL; Mihalczo, John T [ORNL

2007-12-01T23:59:59.000Z

172

Transient analysis for the tajoura critical facility with IRT-2M HEU fuel and IRT-4M leu fuel : ANL independent verification results.  

SciTech Connect

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 one step in the verification process.

Garner, P. L.; Hanan, N. A.

2005-12-02T23:59:59.000Z

173

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

SciTech Connect

Lawrence Livermore National Laboratory (LLNL) uses the LLNL passive-active neutron drum (PAN) shuffler (Canberra Model JCC-92) for accountability measurement of highly enriched uranium (HEU) oxide and HEU in mixed uranium-plutonium (U-Pu) oxide. In June 2002, at the 43rd Annual Meeting of the Institute of Nuclear Material Management, LLNL reported on an extensive effort to calibrate this shuffler, based on standards measurements and extensive simulations, for HEU oxides and mixed U-Pu oxides in thin-walled primary and secondary containers. In August 2002, LLNL began to also use DOE-STD-3013-2000 containers for HEU oxide and mixed U-Pu oxide. These DOE-STD-3013-2000 containers are comprised of a stainless steel convenience can enclosed in welded stainless steel primary and secondary containers. Compared to the double thin-walled containers, the DOE-STD-3013-2000 containers have substantially thicker walls, and the density of materials in these containers was found to extend over a greater range (1.35 g/cm{sup 3} to 4.62 g/cm{sup 3}) than foreseen for the double thin-walled containers. Further, the DOE-STD-3013-2000 Standard allows for oxides containing at least 30 wt% Pu plus U whereas the calibration algorithms for thin-walled containers were derived for virtually pure HEU or mixed U-Pu oxides. An initial series of Monte Carlo simulations of the PAN shuffler response to given quantities of HEU oxide and mixed U-Pu oxide in DOE-STD-3013-2000 containers was generated and compared with the response predicted by the calibration algorithms for thin-walled containers. Results showed a decrease on the order of 10% in the count rate, and hence a decrease in the calculated U mass for measured unknowns, with some varying trends versus U mass. Therefore a decision was made to develop a calibration algorithm for the PAN shuffler unique to the DOE-STD-3013-2000 container. This paper describes that effort and selected unknown item measurement results.

Mount, M E; O' Connell, W J

2005-06-03T23:59:59.000Z

174

Implementation of U.S. transparency monitoring under the U.S./Russian HEU purchase agreement  

SciTech Connect

During the past three years US monitoring at Russian nuclear facilities, subject to the HEU Purchase Agreement, has evolved as MINATOM and DOE negotiators worked to improve transparency rights and as additional Russian facilities began processing HEU. The number of Russian nuclear facilities subject to US monitoring has increased from two in 1996 to the current four. In that time, physical monitoring, which only permitted visual inspections and access to process forms is being supplemented by instrumentation which detects U-235 enrichment of material in containers and instrumentation which is used to confirm that blending of HEU into LEU at the blending facilities is taking place. This paper summarizes the US HEU Transparency monitoring activities performed in Russian facilities. It then summarizes the process used to certify the Blend Down Monitoring System (BDMS) that is currently in use at one of these facilities.

Benton, J B; Glaser, J W; Mastal, E F

1999-07-21T23:59:59.000Z

175

Special Nuclear Materials: EM Manages Plutonium, Highly Enriched Uranium  

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

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

176

Isotope correlation studies relative to high enrichment test reactor fuels  

SciTech Connect

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

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

1978-06-01T23:59:59.000Z

177

Initial report on characterization of excess highly enriched uranium  

SciTech Connect

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

1996-07-01T23:59:59.000Z

178

Profile of World Uranium Enrichment Programs - 2007  

SciTech Connect

It is generally agreed that the most difficult step in building a nuclear weapon is acquiring weapons grade fissile material, either plutonium or highly enriched uranium (HEU). Plutonium is produced in a nuclear reactor, while HEU is produced using a uranium enrichment process. Enrichment is also an important step in the civil nuclear fuel cycle, in producing low enriched uranium (LEU) for use in fuel for nuclear reactors. However, the same equipment used to produce LEU for nuclear fuel can also be used to produce HEU for weapons. Safeguards at an enrichment plant are the array of assurances and verification techniques that ensure uranium is only enriched to LEU, no undeclared LEU is produced, and no uranium is enriched to HEU or secretly diverted. There are several techniques for enriching uranium. The two most prevalent are gaseous diffusion, which uses older technology and requires a lot of energy, and gas centrifuge separation, which uses more advanced technology and is more energy efficient. Gaseous diffusion plants (GDPs) provide about 40% of current world enrichment capacity, but are being phased out as newer gas centrifuge enrichment plants (GCEPs) are constructed. Estimates of current and future enrichment capacity are always approximate, due to the constant upgrades, expansions, and shutdowns occurring at enrichment plants, largely determined by economic interests. Currently, the world enrichment capacity is approximately 53 million kg-separative work units (SWU) per year, with 22 million in gaseous diffusion and 31 million in gas centrifuge plants. Another 23 million SWU/year of capacity are under construction or planned for the near future, almost entirely using gas centrifuge separation. Other less-efficient techniques have also been used in the past, including electromagnetic and aerodynamic separations, but these are considered obsolete, at least from a commercial perspective. Laser isotope separation shows promise as a possible enrichment technique of the future, but has yet to be demonstrated commercially. In the early 1980s, six countries developing gas centrifuge technology (United States, United Kingdom, Germany, the Netherlands, Japan, and Australia) along with the International Atomic Energy Agency (IAEA) and the European Atomic Energy Community (EURATOM) began developing effective safeguards techniques for GCEPs. This effort was known as the Hexapartite Safeguards Project (HSP). The HSP had the goal of maximizing safeguards effectiveness while minimizing the cost to the operator and inspectorate, and adopted several recommendations, such as the acceptance of limited-frequency unannounced access (LFUA) inspections in cascade halls, and the use of nondestructive assay (NDA) measurements and tamper-indicating seals. While only the HSP participants initially committed to implementing all the measures of the approach, it has been used as a model for the safeguards applied to GCEPs in additional states. This report provides a snapshot overview of world enrichment capacity in 2007, including profiles of the uranium enrichment programs of individual states. It is based on open-source information, which is dependent on unclassified sources and may therefore not reflect the most recent developments. In addition, it briefly describes some of the safeguards techniques being used at various enrichment plants, including implementation of HSP recommendations.

Laughter, Mark D [ORNL

2007-11-01T23:59:59.000Z

179

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

Science Conference Proceedings (OSTI)

The U.S. nonproliferation policy 'to minimize, and to the extent possible, eliminate the use of HEU in civil nuclear programs throughout the world' has resulted in the conversion (or scheduled conversion) of many of the U.S. research reactors from high-enriched uranium (HEU) to low-enriched uranium (LEU). A foil fuel appears to offer the best option for using a LEU fuel in the High Flux Isotope Reactor (HFIR) without degrading the performance of the reactor. The purpose of this document is to outline a proposed conceptual fabrication process flow sheet for a new, foil-type, 19.75%-enriched fuel for HFIR. The preparation of the flow sheet allows a better understanding of the costs of infrastructure modifications, operating costs, and implementation schedule issues associated with the fabrication of LEU fuel for HFIR. Preparation of a reference flow sheet is one of the first planning steps needed in the development of a new manufacturing capacity for low enriched fuels for U.S. research and test reactors. The flow sheet can be used to develop a work breakdown structure (WBS), a critical path schedule, and identify development needs. The reference flow sheet presented in this report is specifically for production of LEU foil fuel for the HFIR. The need for an overall reference flow sheet for production of fuel for all High Performance Research Reactors (HPRR) has been identified by the national program office. This report could provide a starting point for the development of such a reference flow sheet for a foil-based fuel for all HPRRs. The reference flow sheet presented is based on processes currently being developed by the national program for the LEU foil fuel when available, processes used historically in the manufacture of other nuclear fuels and materials, and processes used in other manufacturing industries producing a product configuration similar to the form required in manufacturing a foil fuel. The processes in the reference flow sheet are within the bounds of known technology and are adaptable to the high-volume production required to process {approx} 2.5 to 4 tons of U/Mo and produce {approx}16,000 flat plates for U.S. reactors annually ({approx}10,000 of which are needed for HFIR operations). The reference flow sheet is not intended to necessarily represent the best or the most economical way to manufacture a LEU foil fuel for HFIR but simply represents a 'snapshot' in time of technology and is intended to identify the process steps that will likely be required to manufacture a foil fuel. Changes in some of the process steps selected for the reference flow sheet are inevitable; however, no one step or series of steps dominates the overall flow sheet requirements. A result of conceptualizing a reference flow sheet was the identification of the greater number of steps required for a foil process when compared to the dispersion fuel process. Additionally, in most of the foil processing steps, bare uranium must be handled, increasing the complexity of these processing areas relative to current operations. Based on a likely total cost of a few hundred million dollars for a new facility, it is apparent that line item funding will be necessary and could take as much as 8 to 10 years to complete. The infrastructure cost could exceed $100M.

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

2007-09-30T23:59:59.000Z

180

Evaluation of HEU-Beryllium Benchmark Experiments to Improve Computational Analysis of Space Reactors  

SciTech Connect

An assessment was previously performed to evaluate modeling capabilities and quantify preliminary biases and uncertainties associated with the modeling methods and data utilized in designing a nuclear reactor such as a beryllium-reflected, highly-enriched-uranium (HEU)-O2 fission surface power (FSP) system for space nuclear power. The conclusion of the previous study was that current capabilities could preclude the necessity of a cold critical test of the FSP; however, additional testing would reduce uncertainties in the beryllium and uranium cross-section data and the overall uncertainty in the computational models. A series of critical experiments using HEU metal were performed in the 1960s and 1970s in support of criticality safety operations at the Y-12 Plant. Of the hundreds of experiments, three were identified as fast-fission configurations reflected by beryllium metal. These experiments have been evaluated as benchmarks for inclusion in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (IHECSBE). Further evaluation of the benchmark experiments was performed using the sensitivity and uncertainty analysis capabilities of SCALE 6. The data adjustment methods of SCALE 6 have been employed in the validation of an example FSP design model to reduce the uncertainty due to the beryllium cross section data.

John D. Bess; Keith C. Bledsoe; Bradley T. Rearden

2011-02-01T23:59:59.000Z

Note: This page contains sample records for the topic "heu highly enriched" from the National Library of EnergyBeta (NLEBeta).
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181

IAEA Verification Experiment at the Portsmouth Gaseous Diffusion Plant: Report on the Cascade Header Enrichment Monitor  

SciTech Connect

The authors describe the Cascade Header Enrichment Monitor (CHEM) for the Portsmouth Gaseous Diffusion Plant at Piketon, Ohio, and present the calibration and measurement results. The US government has offered excess fissile material that is no longer needed for defense purposes for International Atomic Energy Agency (IAEA) inspection. Measurement results provided by the CHEM were used by the IAEA in a verification experiment to provide confidence that the US successfully blended excess highly enriched uranium (HEU) down to low enriched uranium (LEU). The CHEM measured the uranium enrichment in two cascade header pipes, a 20.32-cm HEU pipe and a 7.62-cm product LEU pipe. The CHEM determines the amount of {sup 235}U from the 185.7-keV gamma-ray photopeak and the amount of total uranium by x-ray fluorescence (XRF) of the 98.4-keV x-ray from uranium with a {sup 57}Co XRF source. The ratio yields the enrichment. The CHEM consists of a collimator assembly, an electromechanically cooled germanium detector, and a rack-mounted personal computer running commercial and custom software. The CHEM was installed in December 1997 and was used by the IAEA inspectors for announced and unannounced inspections on the HEU and LEU header pipes through October 1998. The equipment was sealed with tamper-indicating enclosures when the inspectors were not present.

P. L. Kerr; D. A. Close; W. S. Johnson; R. M. Kandarian; C. E. Moss; C. D. Romero

1999-03-01T23:59:59.000Z

182

Vietnam HEU Removal | National Nuclear Security Administration  

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

commander gives instructions prior to departing in a convoy carrying highly enriched uranium. Facebook Twitter Youtube Flickr Headlines Jul 23, 2013 US, New Zealand...

183

Vietnam HEU Removal | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

employee Igor Bolshinsky measures radioactivity on an ISO container containing highly enriched uranium in Dalat, Vietnam. Facebook Twitter Youtube Flickr Headlines Jul 23, 2013 US,...

184

Vietnam HEU Removal | National Nuclear Security Administration  

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

soldiers receive instructions prior to departing in a convoy carrying highly enriched uranium. Facebook Twitter Youtube Flickr Headlines Jul 23, 2013 US, New Zealand...

185

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

Science Conference Proceedings (OSTI)

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

Robert Gregg; Andrew Worrall

186

Improved criticality search techniques for low and high enriched systems  

SciTech Connect

A new automated search technique has been developed to improve the computational efficiency of performing criticality searches on low and high enriched systems with codes such as ANISN and KENO-IV. The technique employs a least-squares fit to a cubic polynomial on parameter values that have been previously generated either by the Extended Mean Value Theorem (EMVT) or by previous curve fits. The solution of the cubic for its roots at the desired value of K-effective completes one pass for the fixed value search while the solution of its derivative provides information about maximum values. This new search technique has been implemented in a FORTRAN routine called OPTMIZ which will eventually be part of a module in the SCALE system.

Lorek, M.J.; Dodds, H.L.; Petrie, L.M.; Westfall, R.M.

1979-01-01T23:59:59.000Z

187

Validation of NCSSHP for highly enriched uranium systems containing beryllium  

Science Conference Proceedings (OSTI)

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

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

1994-09-29T23:59:59.000Z

188

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

Science Conference Proceedings (OSTI)

Over a period of 30 years, more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited for nuclear data validation and to form the basis for criticality safety benchmarks. A number of the Argonne ZPR/ZPPR critical assemblies have been evaluated as ICSBEP and IRPhEP benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. ZPR-3 Assembly 12 (ZPR-3/12) was designed as a fast reactor physics benchmark experiment with an average core {sup 235}U enrichment of approximately 21 at.%. Approximately 68.9% of the total fissions in this assembly occur above 100 keV, approximately 31.1% occur below 100 keV, and essentially none below 0.625 eV - thus the classification as a 'fast' assembly. This assembly is Fast Reactor Benchmark No. 9 in the Cross Section Evaluation Working Group (CSEWG) Benchmark Specifications and has historically been used as a data validation benchmark assembly. Loading of ZPR-3 Assembly 12 began in late Jan. 1958, and the Assembly 12 program ended in Feb. 1958. The core consisted of highly enriched uranium (HEU) plates, depleted uranium plates and graphite plates loaded into stainless steel drawers which were inserted into the central square stainless steel tubes of a 31 x 31 matrix on a split table machine. The core unit cell consisted of two columns of 0.125 in.-wide (3.175 mm) HEU plates, seven columns of 0.125 in.-wide depleted uranium plates and seven columns of 0.125 in.-wide graphite plates. The length of each column was 9 in. (228.6 mm) in each half of the core. The graphite plates were included to produce a softer neutron spectrum that would be more characteristic of a large power reactor. The axial blanket consisted of 12 in. (304.8 mm) of depleted uranium behind the core. The thickness of the radial blanket was approximately 12 in. and the length of the radial blanket in each half of the matrix was 21 in. (533.4 mm). The assembly geometry approximated a right circular cylinder as closely as the square matrix tubes allowed. According to the logbook and loading records for ZPR-3/12, the reference critical configuration was loading 10 which was critical on Feb. 5, 1958. The subsequent loadings were very similar but less clean for criticality because there were modifications made to accommodate reactor physics measurements other than criticality. Accordingly, ZPR-3/12 loading 10 was selected as the only configuration for this benchmark. As documented below, it was determined to be acceptable as a criticality safety benchmark experiment. An accurate transformation to a simplified model is needed to make any ZPR assembly a practical criticality-safety benchmark. There is simply too much geometric detail in an exact (as-built) model of a ZPR assembly, even a clean core such as ZPR-3/12 loading 10. The transformation must reduce the detail to a practical level without masking any of the important features of the critical experiment. And it must d

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

2010-09-30T23:59:59.000Z

189

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

SciTech Connect

Over a period of 30 years, more than a hundred Zero Power Reactor (ZPR) critical assemblies were constructed at Argonne National Laboratory. The ZPR facilities, ZPR-3, ZPR-6, ZPR-9 and ZPPR, were all fast critical assembly facilities. The ZPR critical assemblies were constructed to support fast reactor development, but data from some of these assemblies are also well suited for nuclear data validation and to form the basis for criticality safety benchmarks. A number of the Argonne ZPR/ZPPR critical assemblies have been evaluated as ICSBEP and IRPhEP benchmarks. Of the three classes of ZPR assemblies, engineering mockups, engineering benchmarks and physics benchmarks, the last group tends to be most useful for criticality safety. Because physics benchmarks were designed to test fast reactor physics data and methods, they were as simple as possible in geometry and composition. The principal fissile species was {sup 235}U or {sup 239}Pu. Fuel enrichments ranged from 9% to 95%. Often there were only one or two main core diluent materials, such as aluminum, graphite, iron, sodium or stainless steel. The cores were reflected (and insulated from room return effects) by one or two layers of materials such as depleted uranium, lead or stainless steel. Despite their more complex nature, a small number of assemblies from the other two classes would make useful criticality safety benchmarks because they have features related to criticality safety issues, such as reflection by soil-like material. ZPR-3 Assembly 6 consisted of six phases, A through F. In each phase a critical configuration was constructed to simulate a very simple shape such as a slab, cylinder or sphere that could be analyzed with the limited analytical tools available in the 1950s. In each case the configuration consisted of a core region of metal plates surrounded by a thick depleted uranium metal reflector. The average compositions of the core configurations were essentially identical in phases A - F. ZPR-3 Assembly 6F (ZPR-3/6F), the final phase of the Assembly 6 program, simulated a spherical core with a thick depleted uranium reflector. ZPR-3/6F was designed as a fast reactor physics benchmark experiment with an average core {sup 235}U enrichment of approximately 47 at.%. Approximately 81.4% of the total fissions in this assembly occur above 100 keV, approximately 18.6% occur below 100 keV, and essentially none below 0.625 eV - thus the classification as a 'fast' assembly. This assembly is Fast Reactor Benchmark No. 7 in the Cross Section Evaluation Working Group (CSEWG) Benchmark Specifications and has historically been used as a data validation benchmark assembly. Loading of ZPR-3/6F began in late December 1956, and the experimental measurements were performed in January 1957. The core consisted of highly enriched uranium (HEU) plates, depleted uranium plates, perforated aluminum plates and stainless steel plates loaded into aluminum drawers, which were inserted into the central square stainless steel tubes of a 31 x 31 matrix on a split table machine. The core unit cell consisted of three columns of 0.125 in.-wide (3.175 mm) HEU plates, three columns of 0.125 in.-wide depleted uranium plates, nine columns of 0.125 in.-wide perforated aluminum plates and one column of stainless steel plates. The maximum length of each column of core material in a drawer was 9 in. (228.6 mm). Because of the goal to produce an approximately spherical core, core fuel and diluent column lengths generally varied between adjacent drawers and frequently within an individual drawer. The axial reflector consisted of depleted uranium plates and blocks loaded in the available space in the front (core) drawers, with the remainder loaded into back drawers behind the front drawers. The radial reflector consisted of blocks of depleted uranium loaded directly into the matrix tubes. The assembly geometry approximated a reflected sphere as closely as the square matrix tubes, the drawers and the shapes of fuel and diluent plates allowed. According to the logbook and loading records for ZPR-3/6F

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

2010-09-30T23:59:59.000Z

190

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

SciTech Connect

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

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

1985-01-01T23:59:59.000Z

191

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

SciTech Connect

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

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

1986-01-01T23:59:59.000Z

192

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

SciTech Connect

DOE has initiated numerous activities to focus on identifying material management strategies to disposition various excess fissile materials. In particular the INEEL has stored 1,700 Kg of offspec HEU at INTEC in CPP-651 vault facility. Currently, the proposed strategies for dispositioning are (a) aqueous dissolution and down blending to LEU via facilities at SRS followed by shipment of the liquid LEU to NFS for fabrication into LWR fuel for the TVA reactors and (b) dilution of the HEU to 0.9% for discard as a waste stream that would no longer have a criticality or proliferation risk without being processed through some type of enrichment system. Dispositioning this inventory as a waste stream via aqueous processing at SRS has been determined to be too costly. Thus, dry blending is the only proposed disposal process for the uranium oxide materials in the CPP-651 vault. Isotopic dilution of HEU to typically less than 20% by dry blending is the key to solving the dispositioning issue (i.e., proliferation) posed by HEU stored at INEEL. RM-2 mill is a technology developed and successfully tested for producing ultra-fine particles by dry grinding. Grinding action in RM-2 mill produces a two million-fold increase in the number of particles being blended in a centrifugal field. In a previous study, the concept of achieving complete and adequate blending and mixing (i.e., no methods were identified to easily separate and concentrate one titanium compound from the other) in remarkably short processing times was successfully tested with surrogate materials (titanium dioxide and titanium mono-oxide) with different particle sizes, hardness and densities. In the current project, the RM-2 milling technology was thoroughly tested with mixtures of natural uranium oxide (NU) and depleted uranium oxide (DU) stock to prove its performance. The effects of mill operating and design variables on the blending of NU/DU oxides were evaluated. First, NU and DU both made of the same oxide, UO{sub 3}, was used in the testing. Next, NU made up of UO{sub 3} and DU made up of UO{sub 2} was used in the test work. In every test, the blend achieved was characterized by spatial sampling of the ground product and analyzing for {sup 235}U concentration. The test work proved that these uranium oxide materials can be blended successfully. The spatial concentration was found to be uniform. Next, sintered thorium oxide pellets were used as surrogate for light water breeder reactor pellets (LWBR). To simulate LWBR pellet dispositioning, the thorium oxide pellets were first ground to a powder form and then the powder was blended with NU. In these tests also the concentration of {sup 235}U and {sup 232}Th in blended products fell within established limits proving the success of RM-2 milling technology. RM-2 milling technology is applicable to any dry radioactive waste, especially brittle solids that can be ground up and mixed with the non-radioactive stock.

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

2003-08-01T23:59:59.000Z

193

Radiation Detection for Active Interrogation of HEU  

SciTech Connect

This report briefly describes the neutrons and gamma rays emitted by active interrogation of HEU, briefly discusses measurement methods, briefly discusses sources and detectors relevant to detection of shielded HEU in Sealand containers, and lists the measurement possibilities for the various sources. All but one of the measurement methods detect radiation emitted by induced fission in the HEU; the exception utilizes nuclear resonance fluorescence. The brief descriptions are supplemented by references. This report presents some active interrogation possibilities but the status of understanding is not advanced enough to select particular methods. Additional research is needed to evaluate these possibilities.

Mihalczo, J.T.

2004-12-09T23:59:59.000Z

194

D&D of the French High Enrichment Gaseous Diffusion Plant  

SciTech Connect

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

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

2003-02-27T23:59:59.000Z

195

RERTR 2009 (Reduced Enrichment for Research and Test Reactors)  

SciTech Connect

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 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.

Totev, T.; Stevens, J.; Kim, Y. S.; Hofman, G.; Matos, J.; Hanan, N.; Garner, P.; Dionne, B.; Olson, A.; Feldman, E.; Dunn, F.; Nuclear Engineering Division; Atomic Research Center; Inst. of Nuclear Physics; LLNL; INL; Korea Atomic Energy Research Inst.; Comisi?n Nacional de Energ?a At?mica; Nuclear Reactor Lab.; Inst. of Atomic Energy-Poland; AECL-Canada; Hungarian Academy of Sciences KFKI Atomic Energy Research Inst.; Japan Atomic Energy Agency; Nuclear Power Inst. of China; Kyoto Univ. Research Reactor Inst.

2010-03-01T23:59:59.000Z

196

Critical masses of highly enriched uranium diluted with matrix material.  

SciTech Connect

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

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

2002-01-01T23:59:59.000Z

197

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

E-Print Network (OSTI)

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

[No author

2006-01-01T23:59:59.000Z

198

Kazakhstan HEU Removal | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

Kazakhstan HEU Removal | National Nuclear Security Administration Kazakhstan HEU Removal | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > content > Four-Year Plan > Kazakhstan HEU Removal Kazakhstan HEU Removal Location Kazakhstan United States 48° 59' 44.1492" N, 67° 3' 37.9692" E See map: Google Maps Printer-friendly version Printer-friendly version

199

Turkey HEU Removal | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

Turkey HEU Removal | National Nuclear Security Administration Turkey HEU Removal | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > content > Four-Year Plan > Turkey HEU Removal Turkey HEU Removal Location Turkey United States 38° 26' 50.2044" N, 40° 15' 14.0616" E See map: Google Maps Printer-friendly version Printer-friendly version

200

Chile HEU Removal | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

HEU Removal Location United States 25 28' 1.4916" S, 69 33' 55.548" W See map: Google Maps Printer-friendly version Printer-friendly version Javascript is required to view...

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


201

France HEU Removal | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

HEU Removal Location United States 45 44' 20.0544" N, 2 17' 6.5616" E See map: Google Maps Printer-friendly version Printer-friendly version Javascript is required to view...

202

Remote Inspection Devices for Spent Reactor Enriched Uranium Fuel Elements  

SciTech Connect

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

Heckendorn, F.M.

2001-01-03T23:59:59.000Z

203

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

SciTech Connect

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

Not Available

1994-01-01T23:59:59.000Z

204

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

SciTech Connect

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

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

1987-01-01T23:59:59.000Z

205

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

Science Conference Proceedings (OSTI)

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

Pesic, Milan P

2003-10-15T23:59:59.000Z

206

Low enrichment fuel conversion for Iowa State University  

SciTech Connect

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

Rohach, A.F.

1992-08-01T23:59:59.000Z

207

Subcritical Neutron Multiplication Measurements of HEU Using Delayed Neutrons as the Driving Source  

SciTech Connect

A new method for the determination of the multiplication of highly enriched uranium systems is presented. The method uses delayed neutrons to drive the HEU system. These delayed neutrons are from fission events induced by a pulsed 14-MeV neutron source. Between pulses, neutrons are detected within a medium efficiency neutron detector using {sup 3}He ionization tubes within polyethylene enclosures. The neutron detection times are recorded relative to the initiation of the 14-MeV neutron pulse, and subsequently analyzed with the Feynman reduced variance method to extract singles, doubles and triples neutron counting rates. Measurements have been made on a set of nested hollow spheres of 93% enriched uranium, with mass values from 3.86 kg to 21.48 kg. The singles, doubles and triples counting rates for each uranium system are compared to calculations from point kinetics models of neutron multiplicity to assign multiplication values. These multiplication values are compared to those from MC NP K-Code calculations.

Hollas, C.L.; Goulding, C.A.; Myers, W.L.

1999-09-20T23:59:59.000Z

208

A confirmatory measurement technique for highly enriched uranium  

SciTech Connect

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

Sprinkle, J.K. Jr.

1987-07-01T23:59:59.000Z

209

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

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

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

210

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

SciTech Connect

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 BY THE FACILITY OPERATOR ON A NEAR-REAL-TIME BASIS INTO A ''MAILBOX'' COMPUTER LOCATED WITHIN A TAMPER-INDICATING ENCLOSURE SEALED BY THE IAEA.

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

2000-07-31T23:59:59.000Z

211

DOE removes all remaining HEU from Hungary | National Nuclear...  

National Nuclear Security Administration (NNSA)

NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > NNSA Blog > DOE removes all remaining HEU from Hungary DOE removes all remaining HEU...

212

US Removes Last Remaining HEU from Czech Republic, Sets Nonproliferati...  

National Nuclear Security Administration (NNSA)

Remaining HEU from Czech Republic, Sets Nonproliferation Milestone | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering...

213

Effect of reduced enrichment on the fuel cycle for research reactors  

SciTech Connect

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

Travelli, A.

1982-01-01T23:59:59.000Z

214

Polyethylene-Reflected Arrays of HEU(93.2) Metal Units Separated by Vermiculite  

SciTech Connect

This benchmark details the results of an experiment performed in the early 1970s as part of a series testing critical configurations in three dimensional arrays. For this experiment, cylinders of 93.2% enriched uranium metal were arranged in a 2x2x2 array inside of a polyethylene reflector. Layers of vermiculite of varying heights were surrounding each cylinder to achieve criticality variations. A total of four experimental configurations were tested by D.W. Magnuson, and detailed in his experimental report “Critical Three-Dimensional Arrays of Neutron Interacting Units: Part IV. Arrays of U(93.2) Metal Reflected by Concrete and Arrays Separated by Vermiculite and Reflected by Polyethylene.” The benchmark HEU-MET-FAST054 is closely related; the results of both experiments are discussed in the same report (Ref. 1) Closely related work has been recorded in HEU-MET-FAST-053, which is a benchmark evaluation of a different series of three dimensional array experiments with four different moderator materials. HEU-MET-FAST-023 and HEU-MET-FAST-026 are also related because they utilize the same metal cylinders as these experiments.

Mackenzie Gorham; J. Blair Briggs; John D. Bess; Virginia Dean; Davis Reed

2010-09-01T23:59:59.000Z

215

Quadrilateral Cooperation on High-density Low-enriched Uranium Fuel  

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

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

216

Quadrilateral Cooperation on High-density Low-enriched Uranium Fuel  

National Nuclear Security Administration (NNSA)

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

217

High-solids enrichment of thermophilic microbial communities and their enzymes on bioenergy feedstocks  

SciTech Connect

Thermophilic microbial communities that are active in a high-solids environment offer great potential for the discovery of industrially relevant enzymes that efficiently deconstruct bioenergy feedstocks. In this study, finished green waste compost was used as an inoculum source to enrich microbial communities and associated enzymes that hydrolyze cellulose and hemicellulose during thermophilic high-solids fermentation of the bioenergy feedstocks switchgrass and corn stover. Methods involving the disruption of enzyme and plant cell wall polysaccharide interactions were developed to recover xylanase and endoglucanase activity from deconstructed solids. Xylanase and endoglucanase activity increased by more than a factor of 5, upon four successive enrichments on switchgrass. Overall, the changes for switchgrass were more pronounced than for corn stover; solids reduction between the first and second enrichments increased by a factor of four for switchgrass while solids reduction remained relatively constant for corn stover. Amplicon pyrosequencing analysis of small-subunit ribosomal RNA genes recovered from enriched samples indicated rapid changes in the microbial communities between the first and second enrichment with the simplified communities achieved by the third enrichment. The results demonstrate a successful approach for enrichment of unique microbial communities and enzymes active in a thermophilic high-solids environment.

Reddy, A. P.; Allgaier, M.; Singer, S.W.; Hazen, T.C.; Simmons, B.A.; Hugenholtz, P.; VanderGheynst, J.S.

2011-04-01T23:59:59.000Z

218

MULTIPLICATION MEASUREMENTS WITH HIGHLY ENRICHED URANIUM METAL SLABS  

SciTech Connect

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

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

1959-07-27T23:59:59.000Z

219

Criticality issues with highly enriched fuels in a repository environment  

SciTech Connect

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

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

1998-03-01T23:59:59.000Z

220

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

DOE Green Energy (OSTI)

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

Lichtenwalter, J.J.

1997-07-01T23:59:59.000Z

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221

Hungary HEU removal | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

removal | National Nuclear Security Administration removal | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > content > Four-Year Plan > Hungary HEU removal Hungary HEU removal Location Hungary United States 47° 11' 51.6336" N, 19° 41' 15" E See map: Google Maps Printer-friendly version Printer-friendly version Javascript is required to view this map.

222

Mexico HEU Removal | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

Removal | National Nuclear Security Administration Removal | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > content > Four-Year Plan > Mexico HEU Removal Mexico HEU Removal Location Mexico United States 24° 24' 35.298" N, 102° 49' 55.3116" W See map: Google Maps Printer-friendly version Printer-friendly version Javascript is required to view this map.

223

Highly enriched multiply-labeled stable isotopic compounds as atmospheric tracers  

DOE Patents (OSTI)

Compounds multiply-labeled with stable isotopes and highly enriched in these isotopes are readily capable of detection in tracer experiments involving high dilutions. Thus, for example, /sup 13/C/sup 18/O/sub 2/ provides a useful tracer for following atmospheric pol lution produced as a result of fossil fuel burning. (Official Gazette)

Goldblatt, M.; McInteer, B.B.

1974-01-29T23:59:59.000Z

224

Dissolved Oralloy standards and the origin of HEU  

SciTech Connect

This report describes an analytical procedure for use in determining the heavy element content of a sample of HEU. Results of the analysis of a specific sample are discussed and some forensic signatures are identified. Two calibrated liquid samples were created, containing known amounts of HEU and contaminants. These samples were counted for gamma rays in the same way that an HEU sample would be treated, and results of the gamma counting are compared with the analytical results.

Moody, K.J.

1994-04-13T23:59:59.000Z

225

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

SciTech Connect

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

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

2001-04-01T23:59:59.000Z

226

Mexico HEU Removal: Fact Sheet | National Nuclear Security Administrat...  

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

Mexico HEU Removal: Fact Sheet | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response...

227

US, International Partners Remove Last Remaining HEU from Vietnam...  

National Nuclear Security Administration (NNSA)

International Partners Remove Last Remaining HEU from Vietnam, Set Nuclear Security Milestone | National Nuclear Security Administration Our Mission Managing the Stockpile...

228

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

SciTech Connect

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

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

1997-09-01T23:59:59.000Z

229

HEU Measurements of Holdup and Recovered Residue in the Deactivation and Decommissioning Activities of the 321-M Reactor Fuel Fabrication Facility at the Savannah River Site  

SciTech Connect

This paper contains a summary of the holdup and material control and accountability (MC&A) assays conducted for the determination of highly enriched uranium (HEU) in the deactivation and decommissioning (D&D) of Building 321-M at the Savannah River Site (SRS). The 321-M facility was the Reactor Fuel Fabrication Facility at SRS and was used to fabricate HEU fuel assemblies, lithium-aluminum target tubes, neptunium assemblies, and miscellaneous components for the SRS production reactors. The facility operated for more than 35 years. During this time thousands of uranium-aluminum-alloy (U-Al) production reactor fuel tubes were produced. After the facility ceased operations in 1995, all of the easily accessible U-Al was removed from the building, and only residual amounts remained. The bulk of this residue was located in the equipment that generated and handled small U-Al particles and in the exhaust systems for this equipment (e.g., Chip compactor, casting furnaces, log saw, lathes A & B, cyclone separator, Freon{trademark} cart, riser crusher, ...etc). The D&D project is likely to represent an important example for D&D activities across SRS and across the Department of Energy weapons complex. The Savannah River National Laboratory was tasked to conduct holdup assays to quantify the amount of HEU on all components removed from the facility prior to placing in solid waste containers. The U-235 holdup in any single component of process equipment must not exceed 50 g in order to meet the container limit. This limit was imposed to meet criticality requirements of the low level solid waste storage vaults. Thus the holdup measurements were used as guidance to determine if further decontamination of equipment was needed to ensure that the quantity of U-235 did not exceed the 50 g limit and to ensure that the waste met the Waste Acceptance Criteria (WAC) of the solid waste storage vaults. Since HEU is an accountable nuclear material, the holdup assays and assays of recovered residue were also important for material control and accountability purposes. In summary, the results of the holdup assays were essential for determining compliance with the Waste Acceptance Criteria, Material Control & Accountability, and to ensure that administrative criticality safety controls were not exceeded. This paper discusses the {gamma}-ray assay measurements conducted and the modeling of the acquired data to obtain measured holdup in process equipment, exhaust components, and fixed geometry scrap cans. It also presents development work required to model new acquisition configurations and to adapt available instrumentation to perform the assays.

DEWBERRY, RAYMOND; SALAYMEH, SALEEM R.; CASELLA, VITO R.; MOORE, FRANK S.

2005-03-11T23:59:59.000Z

230

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

Science Conference Proceedings (OSTI)

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

Primm, Trent [ORNL; Chandler, David [ORNL

2009-01-01T23:59:59.000Z

231

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

SciTech Connect

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

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

2008-06-01T23:59:59.000Z

232

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

Science Conference Proceedings (OSTI)

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

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

2008-02-01T23:59:59.000Z

233

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

Science Conference Proceedings (OSTI)

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

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

2010-09-30T23:59:59.000Z

234

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

SciTech Connect

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

Foley, J.E.

1980-10-01T23:59:59.000Z

235

Mexico HEU Removal: Fact Sheet | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

Mexico HEU Removal: Fact Sheet | National Nuclear Security Administration Mexico HEU Removal: Fact Sheet | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > Media Room > Fact Sheets > Mexico HEU Removal: Fact Sheet Fact Sheet Mexico HEU Removal: Fact Sheet Mar 26, 2012 At the 2012 Nuclear Security Summit, the United States, Mexico and Canada announced the successful removal of HEU from Mexico and conversion of the

236

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

Science Conference Proceedings (OSTI)

The National Ignition Facility (NIF) project, a laser-based Inertial Confinement Fusion (ICF) experiment designed to achieve thermonuclear fusion ignition and burn in the laboratory, is under construction at the Lawrence Livermore National Laboratory (LLNL) and will be completed in April of 2009. Experiments designed to accomplish the NIF's goal will commence in late FY2010 utilizing laser energies of 1 to 1.3 MJ. Fusion yields of the order of 10 to 20 MJ are expected soon thereafter. Laser initiated fusion-fission (LIFE) engines have now been designed to produce nuclear power from natural or depleted uranium without isotopic enrichment, and from spent nuclear fuel from light water reactors without chemical separation into weapons-attractive actinide streams. A point-source of high-energy neutrons produced by laser-generated, thermonuclear fusion within a target is used to achieve ultra-deep burn-up of the fertile or fissile fuel in a sub-critical fission blanket. Fertile fuels including depleted uranium (DU), natural uranium (NatU), spent nuclear fuel (SNF), and thorium (Th) can be used. Fissile fuels such as low-enrichment uranium (LEU), excess weapons plutonium (WG-Pu), and excess highly-enriched uranium (HEU) may be used as well. Based upon preliminary analyses, it is believed that LIFE could help meet worldwide electricity needs in a safe and sustainable manner, while drastically shrinking the nation's and world's stockpile of spent nuclear fuel and excess weapons materials. LIFE takes advantage of the significant advances in laser-based inertial confinement fusion that are taking place at the NIF at LLNL where it is expected that thermonuclear ignition will be achieved in the 2010-2011 timeframe. Starting from as little as 300 to 500 MW of fusion power, a single LIFE engine will be able to generate 2000 to 3000 MWt in steady state for periods of years to decades, depending on the nuclear fuel and engine configuration. Because the fission blanket in a fusion-fission hybrid system is subcritical, a LIFE engine can burn any fertile or fissile nuclear material, including unenriched natural or depleted U and SNF, and can extract a very high percentage of the energy content of its fuel resulting in greatly enhanced energy generation per metric ton of nuclear fuel, as well as nuclear waste forms with vastly reduced concentrations of long-lived actinides. LIFE engines could thus provide the ability to generate vast amounts of electricity while greatly reducing the actinide content of any existing or future nuclear waste and extending the availability of low cost nuclear fuels for several thousand years. LIFE also provides an attractive pathway for burning excess weapons Pu to over 99% FIMA (fission of initial metal atoms) without the need for fabricating or reprocessing mixed oxide fuels (MOX). Because of all of these advantages, LIFE engines offer a pathway toward sustainable and safe nuclear power that significantly mitigates nuclear proliferation concerns and minimizes nuclear waste. An important aspect of a LIFE engine is the fact that there is no need to extract the fission fuel from the fission blanket before it is burned to the desired final level. Except for fuel inspection and maintenance process times, the nuclear fuel is always within the core of the reactor and no weapons-attractive materials are available outside at any point in time. However, an important consideration when discussing proliferation concerns associated with any nuclear fuel cycle is the ease with which reactor fuel can be converted to weapons usable materials, not just when it is extracted as waste, but at any point in the fuel cycle. Although the nuclear fuel remains in the core of the engine until ultra deep actinide burn up is achieved, soon after start up of the engine, once the system breeds up to full power, several tons of fissile material is present in the fission blanket. However, this fissile material is widely dispersed in millions of fuel pebbles, which can be tagged as individual accountable items, and thus made difficult to diver

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

2008-12-23T23:59:59.000Z

237

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

SciTech Connect

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

Tyler Sumner; J. Blair Briggs; Leland Montierth

2007-05-01T23:59:59.000Z

238

TRIGA high wt -% LEU fuel development program. Final report  

SciTech Connect

The principal purpose of this work was to investigate the characteristics of TRIGA fuel where the contained U-235 was in a relatively high weight percent (wt %) of LEU (low enriched uranium - enrichment of less than 20%) rather than a relatively low weight percent of HEU (high enriched uranium). Fuel with up to 45 wt % U was fabricated and found to be acceptable after metallurgical examinations, fission product retention tests and physical property examinations. Design and safety analysis studies also indicated acceptable prompt negative temperature coefficient and core lifetime characteristics for these fuels.

West, G.B.

1980-07-01T23:59:59.000Z

239

Uranium Enrichment  

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

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

240

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

SciTech Connect

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

Crawford, C.; Palmer, B.M.

1992-10-01T23:59:59.000Z

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


241

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

SciTech Connect

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

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

2010-10-01T23:59:59.000Z

242

Critical masses of highly enriched uranium diluted with Gd and polyethylene  

SciTech Connect

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

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

2001-01-01T23:59:59.000Z

243

Overview of reduced-enrichment fuels - development  

SciTech Connect

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

Snelgrove, J.L.

1987-01-01T23:59:59.000Z

244

Isotope Enrichment Detection by Laser Ablation - Dual Tunable Diode Laser Absorption Spectrometry  

Science Conference Proceedings (OSTI)

The rapid global expansion of nuclear energy is motivating the expedited development of new safeguards technology to mitigate potential proliferation threats arising from monitoring gaps within the uranium enrichment process. Current onsite enrichment level monitoring methods are limited by poor sensitivity and accuracy performance. Offsite analysis has better performance, but this approach requires onsite hand sampling followed by time-consuming and costly post analysis. These limitations make it extremely difficult to implement comprehensive safeguards accounting measures that can effectively counter enrichment facility misuse. In addition, uranium enrichment by modern centrifugation leads to a significant proliferation threat, since the centrifuge cascades can quickly produce a significant quantity of highly enriched uranium (HEU). The Pacific Northwest National Laboratory is developing an engineered safeguards approach having continuous aerosol particulate collection and uranium isotope analysis to provide timely detection of HEU production in a low enriched uranium facility. This approach is based on laser vaporization of aerosol particulate samples, followed by wavelength tuned laser diode spectroscopy, to characterize the 235U/238U isotopic ratio by subtle differences in atomic absorption wavelengths arising from differences in each isotope’s nuclear mass, volume, and spin (hyperfine structure for 235U). Environmental sampling media is introduced into a small, reduced pressure chamber, where a focused pulsed laser vaporizes a 10 to 20-µm sample diameter. The ejected plasma forms a plume of atomic vapor. A plume for a sample containing uranium has atoms of the 235U and 238U isotopes present. Tunable diode lasers are directed through the plume to selectively excite each isotope and their presence is detected by monitoring absorbance signals on a shot-to-shot basis. Single-shot detection sensitivity approaching the femtogram range and abundance uncertainty less than 10% have been demonstrated with measurements on surrogate materials. In this paper we present measurement results on samples containing background materials (e.g., dust, minerals, soils) laced with micron-sized target particles having isotopic ratios ranging from 1 to 50%.

Anheier, Norman C.; Bushaw, Bruce A.

2009-07-01T23:59:59.000Z

245

A view from the top: US enrichment Corp. 's William H. Timbers, Jr  

SciTech Connect

Nick Timbers took over as the first Transition Manager of the US Enrichment Corporation upon its founding last July 1st. Although USEC is not involved in negotiating the HEU deal, the fledgling company will be in charge of actually buying and selling the resulting LEU. Whenever the deal is finally signed. After the politics and haggling are over, it will be up to Nick Timbers to make the deal work on the global uranium market. The view from USEC is resolute. No matter what shape the final HEU deal takes, Nick Timbers promises that USEC will remain a competitive supplier of enrichment services. Timbers pledges that any extra costs associated with the HEU deal will not be passed on to customers. He took time out from his recent busy schedule to share his thoughts on HEU and its aftermath.

Giltenan, E.

1993-10-01T23:59:59.000Z

246

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

SciTech Connect

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

Sterbentz, James W

2007-05-01T23:59:59.000Z

247

HEU Minimization and the Reliable Supply of Medical Isotopes Nuclear  

National Nuclear Security Administration (NNSA)

HEU Minimization and the Reliable Supply of Medical Isotopes Nuclear HEU Minimization and the Reliable Supply of Medical Isotopes Nuclear Security Summit: Fact Sheet | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > Media Room > Fact Sheets > HEU Minimization and the Reliable Supply of ... Fact Sheet HEU Minimization and the Reliable Supply of Medical Isotopes Nuclear

248

Czech Republic HEU Removal | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

Czech Republic HEU Removal | National Nuclear Security Administration Czech Republic HEU Removal | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > content > Four-Year Plan > Czech Republic HEU Removal Czech Republic HEU Removal Location Czech Republic United States 49° 35' 23.3628" N, 15° 4' 23.6712" E See map: Google Maps Printer-friendly version Printer-friendly version

249

South Africa HEU Removal | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

South Africa HEU Removal | National Nuclear Security Administration South Africa HEU Removal | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > content > Four-Year Plan > South Africa HEU Removal South Africa HEU Removal Location South Africa United States 30° 33' 35.0604" S, 22° 19' 27.1884" E See map: Google Maps Printer-friendly version Printer-friendly version

250

Fuel cycle cost study with HEU and LEU fuels  

SciTech Connect

Fuel cycle costs are compared for a range of /sup 235/U loadings with HEU and LEU fuels using the IAEA generic 10 MW reactor as an example. If LEU silicide fuels are successfully demonstrated and licensed, the results indicate that total fuel cycle costs can be about the same or lower than those with the HEU fuels that are currently used in most research reactors.

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

1984-01-01T23:59:59.000Z

251

Performance and safety parameters for the high flux isotope reactor  

Science Conference Proceedings (OSTI)

A Monte Carlo depletion model for the High Flux Isotope Reactor (HFIR) Cycle 400 and its use in calculating parameters of relevance to the reactor performance and safety during the reactor cycle are presented in this paper. This depletion model was developed to serve as a reference for the design of a low-enriched uranium (LEU) fuel for an ongoing study to convert HFIR from high-enriched uranium (HEU) to LEU fuel; both HEU and LEU depletion models use the same methodology and ENDF/B-VII nuclear data as discussed in this paper. The calculated HFIR Cycle 400 parameters, which are compared with measurement data from critical experiments performed at HFIR, data included in the HFIR Safety Analysis Report (SAR), or data reported by previous calculations, provide a basis for verification or updating of the corresponding SAR data. (authors)

Ilas, G. [Oak Ridge National Laboratory, 1 Bethel Valley Road, Oak Ridge, TN 37831-6172 (United States); Primm III, T. [Oak Ridge National Laboratory, 1 Bethel Valley Road, Oak Ridge, TN 37831-6172 (United States); Primm Consulting, LLC, 945 Laurel Hill Road, Knoxville, TN 37923 (United States)

2012-07-01T23:59:59.000Z

252

Performance and Safety Parameters for the High Flux Isotope Reactor  

SciTech Connect

A Monte Carlo depletion model for the High Flux Isotope Reactor (HFIR) Cycle 400 and its use in calculating parameters of relevance to the reactor performance and safety during the reactor cycle are presented in this paper. This depletion model was developed to serve as a reference for the design of a low-enriched uranium (LEU) fuel for an ongoing study to convert HFIR from high-enriched uranium (HEU) to LEU fuel; both HEU and LEU depletion models use the same methodology and ENDV/B-VII nuclear data as discussed in this paper. The calculated HFIR Cycle 400 parameters, which are compared when available with measurement data from critical experiments performed at HFIR, data included in the HFIR Safety Analysis Report (SAR), or data reported by previous calculations, provide a basis for verification or updating of the corresponding SAR data.

Ilas, Germina [ORNL; Primm, Trent [Primm Consulting, LLC

2012-01-01T23:59:59.000Z

253

A National Tracking Center for Monitoring Shipments of HEU, MOX, and Spent Nuclear Fuel: How do we implement?  

SciTech Connect

Nuclear material safeguards specialists and instrument developers at US Department of Energy (USDOE) National Laboratories in the United States, sponsored by the National Nuclear Security Administration (NNSA) Office of NA-24, have been developing devices to monitor shipments of UF6 cylinders and other radioactive materials , . Tracking devices are being developed that are capable of monitoring shipments of valuable radioactive materials in real time, using the Global Positioning System (GPS). We envision that such devices will be extremely useful, if not essential, for monitoring the shipment of these important cargoes of nuclear material, including highly-enriched uranium (HEU), mixed plutonium/uranium oxide (MOX), spent nuclear fuel, and, potentially, other large radioactive sources. To ensure nuclear material security and safeguards, it is extremely important to track these materials because they contain so-called “direct-use material” which is material that if diverted and processed could potentially be used to develop clandestine nuclear weapons . Large sources could be used for a dirty bomb also known as a radioactive dispersal device (RDD). For that matter, any interdiction by an adversary regardless of intent demands a rapid response. To make the fullest use of such tracking devices, we propose a National Tracking Center. This paper describes what the attributes of such a center would be and how it could ultimately be the prototype for an International Tracking Center, possibly to be based in Vienna, at the International Atomic Energy Agency (IAEA).

Mark Schanfein

2009-07-01T23:59:59.000Z

254

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

SciTech Connect

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

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

2008-07-15T23:59:59.000Z

255

Progress in converting {sup 99}Mo production from high- to low-enriched uranium--1999.  

SciTech Connect

Over this past year, extraordinary progress has been made in executing our charter to assist in converting Mo-99 production worldwide from HEU to LEU. Building on the successful development of the experimental LEU-foil target, we have designed a new, economical irradiation target. We have also successfully demonstrated, in collaboration with BATAN in Indonesia, that LEU can be substituted for HEU in the Cintichem target without loss of product yield or purity; in fact, conversion may make economic sense. We are interacting with a number of commercial producers--we have begun active collaborations with the CNEA and ANSTO; we are working to define the scope of collaborations with MDS Nordion and Mallinckrodt; and IRE has offered its services to irradiate and test a target at the appropriate time. Conversion of the CNEA process is on schedule. Other papers presented at this meeting will present specific results on the demonstration of the LEU-modified Cintichem process, the development of the new target, and progress in converting the CNEA process.

Snelgrove, J. L.; Vandegrift, G. F.; Conner, C.; Wiencek, T. C.; Hofman, G. L.

1999-09-29T23:59:59.000Z

256

US, International Partners Remove Last Remaining HEU from Vietnam...  

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

and eliminating weapons-usable nuclear materials," said U.S. Secretary of Energy Ernest Moniz. "Today, with the complete removal of all highly enriched uranium from Vietnam, we can...

257

HEU Minimization and the Reliable Supply of Medical Isotopes...  

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

100,000 diagnostic medical procedures globally every day. Today, Mo-99 is produced at aging facilities in Europe, Canada and South Africa primarily using highly-enriched uranium...

258

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

Science Conference Proceedings (OSTI)

An engineering design study that would enable the conversion of the High Flux Isotope Reactor (HFIR) from high-enriched uranium to low-enriched uranium fuel is ongoing at Oak Ridge National Laboratory. The computational models used to search for a low-enriched uranium (LEU) fuel design that would meet the requirements for the conversion study, and the recent results obtained with these models during FY 2009, are documented and discussed in this report. Estimates of relevant reactor performance parameters for the LEU fuel core are presented and compared with the corresponding data for the currently operating high-enriched uranium fuel core. These studies indicate that the LEU fuel design would maintain the current performance of the HFIR with respect to the neutron flux to the central target region, reflector, and beam tube locations.

Ilas, Germina [ORNL; Primm, Trent [ORNL

2009-11-01T23:59:59.000Z

259

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

DOE Patents (OSTI)

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

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

1995-01-01T23:59:59.000Z

260

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

DOE Patents (OSTI)

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

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

1995-05-02T23:59:59.000Z

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


261

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

SciTech Connect

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

1995-12-01T23:59:59.000Z

262

Defining the needs for gas centrifuge enrichment plants advanced safeguards  

Science Conference Proceedings (OSTI)

Current safeguards approaches used by the International Atomic Energy Agency (IAEA) at gas centrifuge enrichment plants (GCEPs) need enhancement in order to verify declared low-enriched (LEU) production, detect undeclared LEU production and detect highly enriched uranium (HEU) production with adequate detection probability using nondestructive 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 UF{sub 6} containers used in the process of enrichment at GCEPs. In verifying declared LEU production, the inspectors also take samples for off-site destructive assay (DA) which provide accurate data, with 0.1% to 0.5% measurement uncertainty, on the enrichment of the UF{sub 6} feed, tails, and product. However, taking samples of UF{sub 6} for off-site analysis is a much more labor and resource intensive exercise for the operator and inspector. Furthermore, the operator must ship the samples off-site to the IAEA laboratory which delays the timeliness of results and interruptions to the continuity of knowledge (CofK) of the samples during their storage and transit. This paper contains an analysis of possible improvements in unattended and attended NDA systems such as process monitoring and possible on-site analysis of DA samples that could reduce the uncertainty of the inspector's measurements and provide more effective and efficient IAEA GCEPs safeguards. We also introduce examples advanced safeguards systems that could be assembled for unattended operation.

Boyer, Brian David [Los Alamos National Laboratory; Erpenbeck, Heather H [Los Alamos National Laboratory; Miller, Karen A [Los Alamos National Laboratory; Swinhoe, Martyn T [Los Alamos National Laboratory; Ianakiev, Kiril [Los Alamos National Laboratory; Marlowe, Johnna B [Los Alamos National Laboratory

2010-01-01T23:59:59.000Z

263

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

SciTech Connect

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

Knight, J.R.

1984-11-01T23:59:59.000Z

264

Methods for Verification of the Hydrogen and Boron Content of the RCSB for Storage of HEU at the HEUMF  

DOE Green Energy (OSTI)

BoroBond{trademark}, which is a ceramic material containing natural boron carbide (B{sub 4}C, a neutron absorber) and water (a neutron attenuator), is the filler material of the Rackable Can Storage Boxes (RCSBs) that will store highly enriched uranium in cans at the Highly Enriched Uranium Materials Facility (HEUMF). Both attenuation and absorption are essential for criticality safety of the fissile material stored in RCSBs. This BoroBond{trademark} material has not yet been used for storage of highly enriched uranium (HEU). To characterize the neutron attenuation and neutron absorption properties of this material, ORNL has performed an extensive series of measurements (over 900) which included: fast neutron and gamma time-of-flight transmission utilizing the Nuclear Materials Identification System (NMIS), thermal and epithermal neutron counting with {sup 3}He proportional counters, and activation analysis with gamma ray spectrometry using a high purity germanium (HPGe) detector. These measurements were performed for a series of 12 x 12-inch square blocks of thickness varying from 2 to 12 inches, with natural B{sub 4}C contents of approximately 0, 2.3, 4.6, and 9 wt%, and varying water contents achieved by baking the blocks to remove approximately 5/6 of the water. These measurements were also performed with a special mockup of the RCSB of BoroBond. material with {approx} 4.6% natural B{sub 4}C. All three methods used Cf-252 sources. This report does not describe these measurements in any detail, but presents a method of verifying and quantifying the B{sub 4}C and hydrogen content of the RCSBs at the factory, upon receipt at Y-12, and at any time later while in use at the HEUMF. The data from these measurements can be used to assess the uniformity of the BoroBond in the RCSB and be stored for future comparisons. The details of these measurements are given in ORNL/TM-2002/254.

Mihalczo, JT

2002-11-21T23:59:59.000Z

265

Radiological consequence analysis with HEU and LEU fuels  

SciTech Connect

A model for estimating the radiological consequences from a hypothetical accident in HEU and LEU fueled research and test reactors is presented. Simple hand calculations based on fission product yield table inventories and non-site specific dispersion data may be adequate in many cases. However, more detailed inventories and site specific data on meteorological conditions and release rates and heights can result in substantial reductions in the dose estimates. LEU fuel gives essentially the same doses as HEU fuel. The plutonium buildup in the LEU fuel does not significantly increase the radiological consequences. The dose to the thyroid is the limiting dose. 10 references, 3 figures, 7 tables.

Woodruff, W.L.; Warinner, D.K.; Matos, J.E.

1984-01-01T23:59:59.000Z

266

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

SciTech Connect

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

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

1986-11-03T23:59:59.000Z

267

PROBING PRE-GALACTIC METAL ENRICHMENT WITH HIGH-REDSHIFT GAMMA-RAY BURSTS  

SciTech Connect

We explore high-redshift gamma-ray bursts (GRBs) as promising tools to probe pre-galactic metal enrichment. We utilize the bright afterglow of a Population III (Pop III) GRB exploding in a primordial dwarf galaxy as a luminous background source, and calculate the strength of metal absorption lines that are imprinted by the first heavy elements in the intergalactic medium (IGM). To derive the GRB absorption line diagnostics, we use an existing highly resolved simulation of the formation of a first galaxy which is characterized by the onset of atomic hydrogen cooling in a halo with virial temperature {approx}> 10{sup 4} K. We explore the unusual circumburst environment inside the systems that hosted Pop III stars, modeling the density evolution with the self-similar solution for a champagne flow. For minihalos close to the cooling threshold, the circumburst density is roughly proportional to (1 + z) with values of about a few cm{sup -3}. In more massive halos, corresponding to the first galaxies, the density may be larger, n {approx}> 100 cm{sup -3}. The resulting afterglow fluxes are weakly dependent on redshift at a fixed observed time, and may be detectable with the James Webb Space Telescope and Very Large Array in the near-IR and radio wavebands, respectively, out to redshift z {approx}> 20. We predict that the maximum of the afterglow emission shifts from near-IR to millimeter bands with peak fluxes from mJy to Jy at different observed times. The metal absorption line signature is expected to be detectable in the near future. GRBs are ideal tools for probing the metal enrichment in the early IGM, due to their high luminosities and featureless power-law spectra. The metals in the first galaxies produced by the first supernova (SN) explosions are likely to reside in low-ionization stages (C II, O I, Si II and Fe II). We show that, if the afterglow can be observed sufficiently early, analysis of the metal lines may distinguish whether the first heavy elements were produced in a pair-instability supernova or a core-collapse (Type II) SN, thus constraining the initial mass function of the first stars.

Wang, F. Y.; Dai, Z. G. [School of Astronomy and Space Science, Nanjing University, Nanjing 210093 (China); Bromm, Volker [Department of Astronomy, University of Texas at Austin, Austin, TX 78712 (United States); Greif, Thomas H. [Max-Planck-Institut fuer Astrophysik, Karl-Schwarzschild-Strasse 1, D-85740 Garching bei Muenchen (Germany); Stacy, Athena [NASA Goddard Space Flight Center, Greenbelt, MD 20771 (United States); Loeb, Abraham [Astronomy Department, Harvard University, 60 Garden St., Cambridge, MA 02138 (United States); Cheng, K. S. [Department of Physics, University of Hong Kong, Pokfulam Road (Hong Kong)

2012-11-20T23:59:59.000Z

268

Analysis of the effectiveness of gas centrifuge enrichment plants advanced safeguards  

SciTech Connect

Current safeguards approaches used by the International Atomic Energy Agency (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 highly enriched uranium (HEU) production with adequate detection 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 235U enrichment of declared UF6 containers used in the process of enrichment at GCEPs. This paper contains an analysis of possible improvements in unattended and attended NDA systems including process monitoring and possible on-site destructive assay (DA) of samples that could reduce the uncertainty of the inspector's measurements. These improvements could reduce the difference between the operator's and inspector's measurements providing more effective and efficient IAEA GCEPs safeguards. We also explore how a few advanced safeguards systems could be assembled for unattended operation. The analysis will focus on how unannounced inspections (UIs), and the concept of information-driven inspections (IDS) can affect probability of detection of the diversion of nuclear materials when coupled to new GCEPs safeguards regimes augmented with unattended systems.

Boyer, Brian David [Los Alamos National Laboratory; Erpenbeck, Heather H [Los Alamos National Laboratory; Miller, Karen A [Los Alamos National Laboratory; Swinjoe, Martyn T [Los Alamos National Laboratory; Ianakiev, Kiril D [Los Alamos National Laboratory; Marlow, Johnna B [Los Alamos National Laboratory

2010-01-01T23:59:59.000Z

269

Measuring progress on HEU minimization - The need for acceleration and addressing 'out-of- scope' activities  

SciTech Connect

Currently, 294 reactors and isotope production facilities use HEU fuel or target material, out of which 154 are used for naval propulsion. These facilities are in annual need of more than 3 500 kg HEU for naval propulsion, more than 900 kg HEU in research reactors, and more than 80 kg HEU for isotope production in civilian facilities, in addition to 6 000 kg HEU in various other types of reactors. 48 civilian research reactors, representing a decrease in the HEU consumption on 278 kg - or 19% compared to the amount of HEU consumed in 1978 in similar facilities, have completed the conversion to LEU as a result over continued international assistance over three decades. The establishment of baseline measurements for assessing the results of the current HEU minimization effort calls for additional focus on the scope and methodology for HEU minimization. The justification for addressing only 54% of the remaining HEU-fueled research reactors as part of the GTRI program should be addressed together with increased focus on facility decommissioning as 120 HEU-fueled reactors with HEU consumption on 450 kg have been shutdown since 1978. There should be no need for converting all the remaining 133 HEU-fueled research reactors as decommissioning and dismantling should play a more prominent role in the future HEU minimization effort. As other sectors reduce the HEU fuel inventory, there is a need to evaluate the risk associated with the continued use of large quantities of weapons-grade HEU fuel for naval propulsion. (author)

Reistad, Ole; Hustveit, Styrkaar [Norwegian Radiation Protection Authority (NRPA), 1332 Osteras (Norway)

2008-07-15T23:59:59.000Z

270

Mixed core conversion study with HEU and LEU fuels  

SciTech Connect

The results of a mixed core study are presented for gradual replacement of HEU fuel with LEU fuel using the IAEA generic 10 MW reactor as an example. The key parameters show that the transition can be accomplished safely and economically.

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

1984-01-01T23:59:59.000Z

271

Derived enriched uranium market  

SciTech Connect

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

Rutkowski, E.

1996-12-01T23:59:59.000Z

272

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

SciTech Connect

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

1994-12-01T23:59:59.000Z

273

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

SciTech Connect

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

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

2004-01-01T23:59:59.000Z

274

Microsoft Word - DOE-ID-INL-11-002.doc  

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

and collocated with R&D and material disposition activities associated with highly enriched uranium (HEU) . This HEU material disposition activity was analyzed in DOEEIS-0240...

275

Transient analyses for HEU and LEU designs of the FRM-II.  

SciTech Connect

An Alternative LEU Design for the FRM-II proposed by the RERTR Program at Argonne National Laboratory (ANL) has a compact core consisting of a single fuel element that uses LEU silicide fuel with a uranium density of 4.5 g/cm{sup 3} and has a power level of 32 MW. Both the HEU design by the Technical University of Munich (TUM) and the alternative LEU design by ANL have the same fuel lifetime (50 days) and the same neutron flux performance (8 x 10{sup 14} n/cm{sup 2}-s in the reflector). LEU silicide fuel with 4.5 g/cm{sup 3} has been thoroughly tested and is fully-qualified, licensable, and available now for use in a high flux reactor such as the FRM-II. Several issues raised by TUM have been addressed in Refs. 1-4. The conclusions of these analyses are summarized below. In this paper, two typical design basis transients are analyzed: control rod withdrawal at different power levels and loss of primary flow. The results show that the HEU and the LEU cores behave in a similar manner and both have excellent safety margins. Based on the excellent results for the Alternative LEU Design that were obtained in all analyses, the RERTR Program reiterates its conclusion that there are no major technical issues regarding use of LEU fuel instead of HEU fuel in the FRM-II and that it is definitely feasible to use LEU fuel in the FRM-II without compromising the safety or performance of the facility.

Hanan, N. A.

1998-10-14T23:59:59.000Z

276

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

SciTech Connect

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

Thomas, W.E.

1976-04-01T23:59:59.000Z

277

Transmutation Analysis of Enriched Uranium and Deep Burn High Temperature Reactors  

Science Conference Proceedings (OSTI)

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

Michael A. Pope

2012-07-01T23:59:59.000Z

278

Fluxes at experiment facilities in HEU and LEU designs for the FRM-II.  

SciTech Connect

An Alternative LEU Design for the FRM-II proposed by the RERTR Program at Argonne National Laboratory (ANL) has a compact core consisting of a single fuel element that uses LEU silicide fuel with a uranium density of 4.5 g/cm{sup 3} and has a power level of 32 MW. Both the HEU design by the Technical University of Munich (TUM) and the alternative LEU design by ANL have the same fuel lifetime(50 days) and the same neutron flux performance (8 x 10{sup 14} n/cm{sup 2}-s in the reflector). LEU silicide fuel with 4.5 g/cm{sup 3} has been thoroughly tested and is fully-qualified, licensable, and available now for use in a high flux reactor such as the FRM-II. Several issues that were raised by TUM have been addressed in Refs. 1-3. The conclusions of these analyses are summarized below. This paper addresses four additional issues that have been raised in several forums, including Ref 4: heat generation in the cold neutron source (CNS), the gamma and fast neutron fluxes which are components of the reactor noise in neutron scattering experiments in the experiment hall of the reactor, a fuel cycle length difference, and the reactivity worth of the beam tubes and other experiment facilities. The results show that: (a) for the same thermal neutron flux, the neutron and gamma heating in the CNS is smaller in the LEU design than in the HEU design, and cold neutron fluxes as good or better than those of the HEU design can be obtained with the LEU desin; (b) the gamma and fast neutron components of the reactor noise in the experiment hall are about the same in both designs; (c) the fuel cycle length is 50 days for both designs; and (d) the absolute value of the reactivity worth of the beam tubes and other experiment facilities is smaller in the LEU design, allowing its fuel cycle length to be increased to 53 or 54 days. Based on the excellent results for the Alternative LEU Design that were obtained in all analyses, the RERTR Program reiterates its conclusion that there are no major technical issues regarding use of LEU fuel instead of HEU fuel in the FRM-II and that it is definitely feasible to use LEU fuel in the FRM-II without compromising the safety or performance of the facility.

Hanan, N. A.

1998-01-16T23:59:59.000Z

279

Field Trial of LANL On-Line Advanced Enrichment Monitor for UF6 GCEP  

Science Conference Proceedings (OSTI)

The outline of this presentation is: (1) Technology basis of on-line enrichment monitoring; (2) Timescale of trial; (3) Description of installed equipment; (4) Photographs; (5) Results; (6) Possible further development; and (7) Conclusions. Summary of the good things about the Advanced Enrichment Monitor (AEM) performance is: (1) High accuracy - normally better than 1% relative, (2) Active system as accurate as passive system, (3) Fast and accurate detection of enrichment changes, (4) Physics is well understood, (5) Elegant method for capturing pressure signal, and (6) Data capture is automatic, low cost and fast. A couple of negative things are: (1) Some jumps in measured passive enrichment - of around +2% relative (due to clock errors?); and (2) Data handling and evaluation is off-line, expensive and very slow. Conclusions are: (1) LANL AEM is being tested on E23 plant at Capenhurst; (2) The trial is going very well; (3) AEM could detect production of HEU at potentially much lower cost than existing CEMO; (4) AEM can measure {sup 235}U assay accurately; (5) Active system using X-Ray source would avoid need for pressure measurement; (6) Substantial work lies ahead to go from current prototype to a production instrument.

Ianakiev, Kiril D. [Los Alamos National Laboratory; Lombardi, Marcie [Los Alamos National Laboratory; MacArthur, Duncan W. [Los Alamos National Laboratory; Parker, Robert F. [Los Alamos National Laboratory; Smith, Morag K. [Los Alamos National Laboratory; Keller, Clifford [Los Alamos National Laboratory; Friend, Peter [URENCO; Dunford, Andrew [URENCO

2012-07-13T23:59:59.000Z

280

Final report on the project entitled: Highly Preheated Combustion Air System with/without Oxygen Enrichment for Metal Processing Furnaces  

SciTech Connect

This work develops and demonstrates a laboratory-scale high temperature natural gas furnace that can operate with/without oxygen enrichment to significantly improve energy efficiency and reduce emissions. The laboratory-scale is 5ft in diameter & 8ft tall. This furnace was constructed and tested. This report demonstrates the efficiency and pollutant prevention capabilities of this test furnace. The project also developed optical detection technology to control the furnace output.

Arvind Atreya

2007-02-16T23:59:59.000Z

Note: This page contains sample records for the topic "heu highly enriched" from the National Library of EnergyBeta (NLEBeta).
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281

NNSA, Republic of Korea Ministry Agree to Minimize Use of HEU...  

National Nuclear Security Administration (NNSA)

Republic of Korea Ministry Agree to Minimize Use of HEU in Nuclear Reactors | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation...

282

NNSA Works to Minimize the use of HEU in Medical Isotope Production...  

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

Works to Minimize the use of HEU in Medical Isotope Production | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the...

283

SIMULATIONS FOR ACTIVE INTERROGATION OF HEU IN CARGO CONTAINERS  

DOE Green Energy (OSTI)

We describe the results of a Monte Carlo simulation 10 investigate the feasibility of using a pulsed deuterium-tritium (D-T) neutron technique for active interrogation of special nuclear material in cargo containers. Time distributions of fission neutrons from highly enriched uranium induced by a pulsed D-T neutron source were calculated for cargo containers with different hydrogen contents. A simple detector system with polyethylene and cadmium was modeled to calculate the two-group neutron flux at the detector.

LEE, SANG Y. [Los Alamos National Laboratory; BEDDINGFIELD, DAVID H. [Los Alamos National Laboratory; PARK, JAEYOUNG [Los Alamos National Laboratory

2007-01-22T23:59:59.000Z

284

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

SciTech Connect

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

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

1986-11-01T23:59:59.000Z

285

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

SciTech Connect

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

Adams, T.M.

1998-11-25T23:59:59.000Z

286

Design, construction, and operation of a laboratory scale reactorfor the production of high-purity, isotopically enriched bulksilicon  

DOE Green Energy (OSTI)

The design and operation of a recirculating flow reactor designed to convert isotopically enriched silane to polycrystalline Si with high efficiency and chemical purity is described. The starting material is SiF{sub 4}, which is enriched in the desired isotope by a centrifuge method and subsequently converted to silane. In the reactor, the silane is decomposed to silicon on the surface of a graphite starter rod (3 mm diameter) heated to 700-750 C. Flow and gas composition (0.3-0.5% silane in hydrogen) are chosen to minimize the generation of particles by homogeneous nucleation of silane and to attain uniform deposition along the length of the rod. Growth rates are 5 {micro}m/min, and the conversion efficiency is greater than 95%. A typical run produces 35 gm of polycrystalline Si deposited along a 150 mm length of the rod. After removal of the starter rod, dislocation-free single crystals are formed by the floating zone method. Crystals enriched in all 3 stable isotopes of Si have been made: {sup 28}Si (99.92%), {sup 29}Si (91.37%), and {sup 30}Si (88.25%). Concentrations of electrically active impurities (P and B) are as low as mid-10{sup 13} cm{sup -3}. Concentrations of C and O lie below 10{sup 16} and 10{sup 15} cm{sup -3}, respectively.

Ager III, J.W.; Beeman, J.W.; Hansen, W.L.; Haller, E.E.

2004-12-20T23:59:59.000Z

287

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

SciTech Connect

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

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

1993-06-04T23:59:59.000Z

288

Development of CFD models to support LEU Conversion of ORNL s High Flux Isotope Reactor  

SciTech Connect

The US Department of Energy s National Nuclear Security Administration (NNSA) is participating in the Global Threat Reduction Initiative to reduce and protect vulnerable nuclear and radiological materials located at civilian sites worldwide. As an integral part of one of NNSA s subprograms, Reduced Enrichment for Research and Test Reactors, HFIR is being converted from the present HEU core to a low enriched uranium (LEU) core with less than 20% of U-235 by weight. Because of HFIR s importance for condensed matter research in the United States, its conversion to a high-density, U-Mo-based, LEU fuel should not significantly impact its existing performance. Furthermore, cost and availability considerations suggest making only minimal changes to the overall HFIR facility. Therefore, the goal of this conversion program is only to substitute LEU for the fuel type in the existing fuel plate design, retaining the same number of fuel plates, with the same physical dimensions, as in the current HFIR HEU core. Because LEU-specific testing and experiments will be limited, COMSOL Multiphysics was chosen to provide the needed simulation capability to validate against the HEU design data and previous calculations, and predict the performance of the proposed LEU fuel for design and safety analyses. To achieve it, advanced COMSOL-based multiphysics simulations, including computational fluid dynamics (CFD), are being developed to capture the turbulent flows and associated heat transfer in fine detail and to improve predictive accuracy [2].

Khane, Vaibhav B [ORNL; Jain, Prashant K [ORNL; Freels, James D [ORNL

2012-01-01T23:59:59.000Z

289

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

Science Conference Proceedings (OSTI)

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

Ilas, Germina [ORNL; Primm, Trent [ORNL

2010-01-01T23:59:59.000Z

290

Removal of Last Remaining HEU from Vietnam - Time Lapse Video | National  

National Nuclear Security Administration (NNSA)

Removal of Last Remaining HEU from Vietnam - Time Lapse Video | National Removal of Last Remaining HEU from Vietnam - Time Lapse Video | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > Media Room > Video Gallery > Removal of Last Remaining HEU from Vietnam ... Removal of Last Remaining HEU from Vietnam - Time Lapse Video Removal of Last Remaining HEU from Vietnam - Time Lapse Video

291

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

SciTech Connect

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

Cantrell, J.

2012-05-23T23:59:59.000Z

292

Abundances at High Redshifts: the Chemical Enrichment History of Damped Lyman-alpha Galaxies  

E-Print Network (OSTI)

Damped Lyman-alpha absorption systems found in the spectra of high redshift quasars are believed to trace the interstellar gas in high redshift galaxies. In this paper, we study the elemental abundances of C, N, O, Al, Si, S, Cr, Mn, Fe, Ni, and Zn in a sample of 14 damped Lyman-alpha systems using high quality echelle spectra of quasars obtained with the 10m Keck telescope. These abundances are combined with similar measurements in the literature in order to investigate the chemical evolution of damped Lyman-alpha galaxies in the redshift range 0.7nature of the star formation process in damped Lyman-alpha galaxies, and the nature of damped Lyman-alpha galaxies themselves.

Limin Lu; Wallace L. W. Sargent; Thomas A. Barlow

1996-06-07T23:59:59.000Z

293

NNSA Awards Funding to Accelerate Non-HEU-Based Production of Molybdenum-99  

National Nuclear Security Administration (NNSA)

Funding to Accelerate Non-HEU-Based Production of Molybdenum-99 Funding to Accelerate Non-HEU-Based Production of Molybdenum-99 in the United States | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > Media Room > Press Releases > NNSA Awards Funding to Accelerate Non-HEU-Based Production ... Press Release NNSA Awards Funding to Accelerate Non-HEU-Based Production of Molybdenum-99

294

Mapping Metal-Enriched High Velocity Clouds to Very Low HI Column Densities  

E-Print Network (OSTI)

Our galaxy is the nearest known quasar absorption line system, and it uniquely provides us with an opportunity to probe multiple lines of sight through the same galaxy. This is essential for our interpretations of the complex kinematic profiles seen in the MgII absorption due to lines of sight through intermediate redshift galaxies. The Milky Way halo has never been probed for high velocity clouds below the 21-cm detection threshold of N(HI)~10^18 cm-2. Through a survey of MgII absorption looking toward the brightest AGNs and quasars, it will be possible to reach down a few orders of magnitude in HI column density. The analogs to the high velocity components of the MgII absorption profiles due to intermediate redshift galaxies should be seen. We describe a program we are undertaking, and present some preliminary findings.

Chris Churchill; Jane Charlton; Joe Masiero

2001-08-13T23:59:59.000Z

295

Thermal breeder fuel enrichment zoning  

DOE Patents (OSTI)

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

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

1992-01-01T23:59:59.000Z

296

Biosynthetic Approaches to Isotope Enrichment for Applications in Neutron Scattering and High Field NMR Spectroscopy: Methylotrophic  

DOE Green Energy (OSTI)

Limitations in current isotopic labeling methods present a substantial bottleneck for the application of advanced structural techniques to many important biochemical problems. New tools are required to efficiently produce the necessary labeling patterns in biochemical precursors and incorporate them into protein molecules for structural studies. This project proposed involved one aspect of this problem, the development of expression vectors for a methylotrophic bacterium, Methylobacterium extorquens AM1. If high-level, efficient expression could be obtained in such a bacterium, it would be possible to use low-cost {sup 2}H- and/or {sup 13}C-labeled substrates such as methanol to label proteins. The Lidstrom laboratory at the University of Washington worked closely with the collaborators at Los Alamos National Laboratories in the development and use of these vectors. (1) Overexpression of a target gene, bacterial dehalogenase--This enzyme was expressed in Methylobacterium extorquens AM1 using a high level methanol-inducible promoter, the mxaF promoter. High expression was achieved, but most was in an insoluble form. They expressed this protein in a mutant lacking polybetahydroxybutyrate granules, and high expression was achieved, up to 10% of the total soluble protein. The recombinant protein was purified and shown to be active, with characteristics similar to the enzyme produced in E. coli. (2) Development of regulated expression systems--A number of regulated promoters were tested in M. extorquens AM1, the most promising of which appeared to be the E. coli lac promoter coupled to the Laciq regulator. The repressor was shown to be active and a chromosomal insertion construct was generated that repressed the low-level lac promoter activity in M. extorquens AM1. However, IPTG induced this system only poorly. A number of studies were carried out leading to the conclusion that IPTG entered the cell but was exported by one or more export pumps. Target genes for such pumps were mutated but none of these showed increased induction. A number of methods were used to permeabilize the cell, and a 2-fold increase in induction was obtained with one of these. The activity of the lac promoter was increased by inserting a recently-identified M. extorquens AM1 enhancer element upstream. The promoter increased in activity 5-6 fold with this addition. In summary, they have developed a suite of expression tools and host mutant strains for expressing a variety of heterologous proteins in this methylotroph. These are now available for testing by the LANL collaborators in labeling reactors to obtain labeled proteins of interest.

Mary E. lidstrom

2004-09-15T23:59:59.000Z

298

EPA Update: NESHAP Uranium Activities  

E-Print Network (OSTI)

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

299

T&E Protocol N42.43, 2009  

Science Conference Proceedings (OSTI)

... Special Nuclear Materials: Uranium ... Highly Enriched Uranium (HEU), Reactor Grade Plutonium ... Radiation Monitors Used for Homeland Security ...

2011-09-19T23:59:59.000Z

300

T&E Protocol N42.48, 2009  

Science Conference Proceedings (OSTI)

... Special nuclear materials: HEU (highly enriched ... 235U >90%), Pu [Reactor grade plutonium ... Radiation Detectors (SPRDs) for Homeland Security. ...

2011-09-19T23:59:59.000Z

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


301

QU Car: a very high luminosity nova-like binary with a carbon-enriched companion  

E-Print Network (OSTI)

QU Car is listed in cataclysmic variable star catalogues as a nova-like variable. This little-studied, yet bright interacting binary is re-appraised here in the light of new high-quality ultraviolet (UV) interstellar line data obtained with STIS on board the Hubble Space Telescope. The detection of a component of interstellar absorption at a mean LSR velocity of $-$14 km s$^{-1}$ indicates that the distance to QU Car may be $\\sim$2 kpc or more -- a considerable increase on the previous lower-limiting distance of 500 pc. If so, the bolometric luminosity of QU Car could exceed $10^{37}$ ergs s$^{-1}$. This would place this binary in the luminosity domain occupied by known compact-binary supersoft X-ray sources. Even at a 500 pc, QU Car appears to be the most luminous nova-like variable known. New intermediate dispersion optical spectroscopy of QU Car spanning 3800--7000 \\AA is presented. These data yield the discovery that C{\\sc iv} $\\lambda\\lambda$5801,12 is present as an unusually prominent emission line in an otherwise low-contrast line spectrum. Using measurements of this and other lines in a recombination line analysis, it is shown that the C/He abundance as proxied by the n(C$^{4+}$)/n(He$^{2+}$) ratio may be as high as 0.06 (an order of magnitude higher than the solar ratio). Furthermore, the C/O abundance ratio is estimated to be greater than 1. These findings suggest that the companion in QU Car is a carbon star. If so, it would be the first example of a carbon star in such a binary. An early-type R star best matches the required abundance pattern and could escape detection at optical wavelengths provided the distance to QU Car is $\\sim$2 kpc or more.

J. E. Drew; L. E. Hartley; K. S. Long; J. van der Walt

2002-09-16T23:59:59.000Z

302

A Robust and Flexible Design for GCEP Unattended Online Enrichment Monitoring: An OLEM Collection Node Network  

SciTech Connect

Oak Ridge National Laoratory (ORNL) has engineered an on-line enrichment monitor (OLEM) to continuously measure U-235 emissions from the UF6 gas flowing through a unit header pipe of a gas centrifuge enrichment plant (GCEP) as a component of the International Atomic Energy Agency s (IAEA) new generation of technology to support enrichment plant safeguards1. In contrast to other enrichment monitoring approaches, OLEM calibrates and corrects for the pressure and temperature dependent UF6 gas-density without external radiation sources by using the inherent unit header pipe pressure dynamics and combining U-235 gamma-ray spectrometery using a shielded NaI detector with gas pressure and temperature data near the spectrum measurement point to obtain the enrichment of the gas as a function of time. From a safeguards perspective, OLEM can provide early detection of a GCEP being misused for production of highly enriched uranium, but would not detect directly the isolation and use of a cascade within the production unit to produce HEU. OLEM may also reduce the number of samples collected for destructive assay and, if coupled with load cell monitoring, could support isotope mass balance verification and unattended cylinder verification. The earlier paper presented OLEM as one component along with shared load cells and unattended cylinder verification, in the IAEA emering toolbox for unattended instruments at GCEPs1 and described the OLEM concept and how previous modeling studies and field measurements helped confirm the viability of a passive on-line enrichment monitor for meeting IAEA objectives and to support the development of performance targets. Phase I of the United States Support Program (USSP) OLEM project completed a preliminary hardware, software and communications design; phase II will build and test field prototypes in controlled laboratory settings and then at an operational facility. That paper also discussed many of the OLEM collection node commercial off the shelf (COTS) components and summarized the OLEM collection node data security provisions. This paper will discuss a secure and redundant network of OLEM collection nodes, auxiliary detection units and supporting junction boxes distributed throughout a facility for monitoring enrichment on product, feed and tails unit header pipes; the purpose and capability of the built-in Electronic Optical Sealing System (EOSS) network gateway; and a network approach for obtaining reliable and authenticated pressure measurements.

Younkin, James R [ORNL; March-Leuba, Jose A [ORNL; Garner, James R [ORNL

2013-01-01T23:59:59.000Z

303

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

SciTech Connect

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

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

2004-02-01T23:59:59.000Z

304

NMIS with Imaging and Gamma Ray Spectrometry for Pu, HEU, HE and Other Materials  

SciTech Connect

The Nuclear Material Identification System (NMIS) has been under development at ORNL and the National Nuclear Security Administration (NNSA) Y-12 National Security Complex since 1984. In the mid-1990s, what is now the US Department of Energy (DOE) Office of Nuclear Verification (ONV) realized that it was a useful technology for future arms control treaty applications and supported further development of the system. In 2004, fast-neutron imaging was incorporated into the system. In 2007, the ONV decided to develop a fieldable version of the system, designated as FNMIS, for potential use in future treaties. The FNMIS is being developed to be compatible with the eventual incorporation of gamma-ray spectrometry and an information barrier. This report addresses how and what attributes could be determined by the FNMIS system with gamma-ray spectrometry. The NMIS is a time-dependent coincidence system that incorporates tomographic imaging (including mapping of the fission sites) and gamma-ray spectrometry. It utilizes a small, lightweight (30 lb), portable deuterium-tritium (DT) neutron (14.1 MeV) generator (4 x 10{sup 7} neutrons/second) for active interrogation and can also perform passive interrogation. A high-purity germanium (HPGe) gamma-ray detector with multichannel analysis can be utilized in conjunction with the source for active interrogation or passively. The system uses proton recoil scintillators: 32 small 2.5 x 2.5 x 10.2-cm-thick plastic scintillators for imaging and at least two 2 x 2 arrays of 27 x 27 x 10-cm-thick plastic scintillators that detect induced fission radiation. The DT generator contains an alpha detector that time and directionally tags a fan beam of some of the neutrons emitted and subdivides it into pixels. A fast (1 GHz) time correlation processor measures the time-dependent coincidence among all detectors in the system. A computer-controlled scanner moves the small detectors and the source appropriately for scanning a target object for imaging. The system is based on detection of transmitted 14.1 MeV neutrons, fission neutrons, and gamma rays from spontaneous, inherent source fission of the target, fission neutrons and gamma rays induced by the external DT source, gamma rays from natural emissions of uranium and plutonium, and induced gamma-ray emission by the interaction of the 14.1 MeV neutrons from the DT source. The NMIS can and has been used with a time-tagged californium spontaneous fission source. It has also been used with pulsed interrogation sources such as LINACs, DT, and deuterium-deuterium (DD) sources. This system is uniquely suited for detection of shielded highly enriched uranium (HEU), plutonium, and other special nuclear materials and detection of high explosives (HE) and chemical agents. The NMIS will be adapted to utilize a trusted processor that incorporates information barrier and authentication techniques using open software and then be useful in some international applications for materials whose characteristics may be classified. The proposed information barrier version of the NMIS system would consist of detectors and cables, the red (classified side) computer system, which processes the data, and the black (unclassified side) computer, which handles the computer interface. The system could use the 'IB wrapper' concept proposed by Los Alamos National Laboratory and the software integrity (digital signatures) system proposed by Sandia. Since it is based entirely on commercially available components, the entire system, including NMIS data acquisition boards, can be built with commercial off-the-shelf components. This system is being developed into a fieldable system (FNMIS) for potential arms control treaties by the ONV. The system will be modularly constructed with the RF shielded modules connected to the processor by appropriate control and signal cable in metal conduit. The FNMIS is presently being designed for eventual incorporation of gamma-ray spectrometry and an information barrier to protect classified information. The system hardware and software can be configu

Mihalczo, John T [ORNL; Mullens, James Allen [ORNL

2012-03-01T23:59:59.000Z

305

HEU Holdup Measurements in 321-M A-Lathe  

SciTech Connect

The Analytical Development Section of SRTC 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.

Dewberry, R.A.

2001-09-18T23:59:59.000Z

306

Studies of Past Operations at the High Flux Isotope Reactor  

Science Conference Proceedings (OSTI)

During the past year, two topics related to past operations of the High Flux Isotope Reactor (HFIR) were reviewed in response to on-going programs at Oak Ridge National Laboratory (ORNL). Currently, studies are being conducted to determine if HFIR can be converted from high enriched uranium (HEU) fuel to low enriched uranium (LEU). While the basis for conversion is the current performance of the reactor, redesign studies revealed an apparent slight degradation in performance of the reactor over its 40 year lifetime. A second program requiring data from HFIR staff is the Integrated Facility Disposition Project (IFDP). The IFDP is a program that integrates environmental cleanup with modernization and site revitalization plans and projects. Before a path of disposal can be established for discharged HFIR beryllium reflector regions, the reflector components must be classified as to type of waste and specifically, determine if they are transuranic waste.

Chandler, David [ORNL; Primm, Trent [ORNL

2009-01-01T23:59:59.000Z

307

Analyses of Greek Research Reactor with mixed HEU-LEU Be reflected core  

SciTech Connect

The fuel-cycle analyses presented in this paper provide specific steps to be taken in the transition from a 36-element water-reflected HEU core to a 33-element LEU equilibrium core with a Be reflector on two faces. The first step will be to install the Be reflector and remove the highest burnup HEU fuel. The smaller Be-reflected core will be refueled with LEU fuel. All analyses were performed using a planar 5-group REBUS3 model benchmarked to VIM Monte Carlo. In addition to fuel cycle results, the control rod worth, reactivity response to increased fuel and water temperature and decreased water density were compared for the transition core and the reference HEU core.

Deen, J.R.; Snelgrove, J.L. [Argonne National Lab., IL (United States); Papastergiou, K. [National Center for Scientific Research, Athens (Greece)

1993-12-31T23:59:59.000Z

308

Comparison of safety parameters and transient behavior of a generic 10 MW reactor with HEU and LEU fuels  

SciTech Connect

Key safety parameters are compared for equilibrium cores of the IAEA generic 10 MW reactor with HEU and LEU fuels. These parameters include kinetics parameters, reactivity feedback coefficients, control rod worths, power peaking factors, and shutdown margins. Reactivity insertion and loss-of-flow transients are compared. Results indicate that HEU and LEU cores will behave in a very similar manner.

Matos, J.E.; Freese, K.E.; Woodruff, W.L.

1983-01-01T23:59:59.000Z

309

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

SciTech Connect

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

Senn, R.L.

1989-04-01T23:59:59.000Z

310

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

SciTech Connect

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

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

2006-02-01T23:59:59.000Z

311

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

SciTech Connect

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

Michael Tyacke; Dr. Igor Bolshinsky; Jeff Chamberlin

2008-07-01T23:59:59.000Z

312

Nuclear Fuels  

Science Conference Proceedings (OSTI)

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

313

Development and Irradiation Performance of U-Mo Monolithic Fuel ...  

Science Conference Proceedings (OSTI)

... that will enable the conversion of research and test reactors from high enriched uranium (HEU) to low enriched uranium (LEU) without loss of performance.

314

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

SciTech Connect

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

Travelli, A.

1994-05-01T23:59:59.000Z

315

Recent Studies Related to Past Operations at the High Flux Isotope Reactor  

Science Conference Proceedings (OSTI)

During the past year, two topics related to past operations of the High Flux Isotope Reactor (HFIR) were reviewed in response to on-going programs at Oak Ridge National Laboratory (ORNL). Currently, studies are being conducted to determine if HFIR can be converted from high enriched uranium (HEU) fuel to low enriched uranium (LEU). While the basis for conversion is the current performance of the reactor, redesign studies revealed an apparent slight degradation in performance of the reactor over its 40 year lifetime. A second program requiring data from HFIR staff is the Integrated Facility Disposition Project (IFDP). The IFDP is a program that integrates environmental cleanup with modernization and site revitalization plans and projects. Before a path of disposal can be established for discharged HFIR beryllium reflector regions, the reflector components must be classified as to type of waste and specifically, determine if they are transuranic waste.

Chandler, David [ORNL; Primm, Trent [ORNL

2009-01-01T23:59:59.000Z

316

New generation enrichment monitoring technology for gas centrifuge enrichment plants  

SciTech Connect

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

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

2008-01-01T23:59:59.000Z

317

Results of transient /accident analysis for the HEU, first mixed HEU-LEU and for the first full LEU cores of the WWR-SM reactor at INP AS RUZ  

SciTech Connect

The WWR-SM reactor in Uzbekistan is preparing for the conversion from HEU (36%) fuel to LEU (19.8%) fuel. During this conversion, the HEU fuel assemblies (IRT-3M FA) being discharged at the end of each cycle will be replaced by LEU fuel assemblies (IRT-4M FA); this gradual conversion requires 9 cycles. The safety analysis report for this conversion process has been prepared. This paper presents selected results for postulated transient/accidents during this conversion process; results for transient analysis for the HEU core, the 1st mixed (HEU-LEU) core, and for the first full LEU core are presented for the following initiators: control rod motion (2 cases), loss of power, and FA blockage. These results show that safety is maintained for all transients analyzed and that the behavior of all the analyzed cores is essentially the same. (author)

Baytelesov, S.A.; Dosimbaev, A.A.; Kungurov, F.R.; Salikhbaev, U.S. [Institute of Nuclear Physics, Ulugbek, 100214 Tashkent (Uzbekistan)

2008-07-15T23:59:59.000Z

318

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

SciTech Connect

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

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

1995-07-01T23:59:59.000Z

319

A detailed neutronics comparison of the university of Florida training reactor (UFTR) current HEU and proposed LEU cores  

SciTech Connect

For over 35 years, the UFTR highly-enriched core has been safely operated. As part of the Reduced Enrichment for Research and Test Reactors Program, the core is currently being converted to low-enriched uranium fuel. The analyses presented in this paper were performed to verify that, from a neutronic perspective, a proposed low-enriched core can be operated as safely and as effectively as the highly-enriched core. Detailed Monte Carlo criticality calculations are performed to determine: i) Excess reactivity for different core configurations, ii) Individual integral blade worth and shutdown margin, iii) Reactivity coefficients and kinetic parameters, and iv) Flux profiles and core six-factor formula parameters. (authors)

Dionne, B.; Haghighat, A.; Yi, C.; Smith, R.; Ghita, G.; Manalo, K.; Sjoden, G.; Huh, J.; Baciak, J.; Mock, T.; Wenner, M. [Dept. of Nuclear and Radiological Engineering, Univ. of Florida, Gainesville, FL (United States); Matos, J.; Stillman, J. [Reduced Enrichment for Research and Test Reactors Program, Argonne National Laboratory, Argonne, IL (United States)

2006-07-01T23:59:59.000Z

320

NNSA Reaches LEU Disposal Milestone | National Nuclear Security...  

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

HEU Blend Down Project, which downblends surplus U.S. weapons-grade highly enriched uranium (HEU), located at SRS and other DOE and NNSA sites, into LEU for peaceful use...

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


321

Performance and fuel cycle cost study of the R2 reactor with HEU and LEU fuels  

SciTech Connect

A systematic study of the experiment performance and fuel cycle costs of the 50 MW R2 reactor operated by Studsvik Energiteknik AB has been performed using the current R2 HEU fuel, a variety of LEU fuel element designs, and two core-box/reflector configurations. The results include the relative performance of both in-core and ex-core experiments, control rod worths, and relative annual fuel cycle costs.

Pond, R.B.; Freese, K.E.; Matos, J.E.

1984-01-01T23:59:59.000Z

322

Foreign research reactor irradiated nuclear fuel inventories containing HEU and LEU of United States origin  

SciTech Connect

This report provides estimates of foreign research reactor inventories of aluminum-based and TRIGA irradiated nuclear fuel elements containing highly enriched and low enriched uranium of United States origin that are anticipated in January 1996, January 2001, and January 2006. These fuels from 104 research reactors in 41 countries are the same aluminum-based and TRIGA fuels that were eligible for receipt under the Department of Energy`s Offsite Fuels Policy that was in effect in 1988. All fuel inventory and reactor data that were available as of December 1, 1994, have been included in the estimates of approximately 14,300 irradiated fuel elements in January 1996, 18,800 in January 2001, and 22,700 in January 2006.

Matos, J.E.

1994-12-01T23:59:59.000Z

323

Advancing Methods for Determining the Source of HEU Used in Terrorist Nuclear Weapon  

E-Print Network (OSTI)

An algorithm was developed that uses measured isotopic ratios from fission product residue following the detonation of a high-enriched uranium nuclear weapon to compute the original attributes of the material used in the device. The specific attributes assessed are the uranium isotopics (considering 234U, 235U, 236U, and 238U) and the enrichment process used to create the material (e.g., gaseous diffusion, gas centrifuge, etc.). Using the original attributes of the weapon significantly increases the probability of identifying the perpetrator of the attack. In this study, research was conducted to perform sensitivity analysis of the calculated values, analyze alternate enrichment methods, determine the source (uranium mine) from which the feed material was taken and assess potential “spoofing” techniques. The purpose of this research was to verify that the analytical method developed would remain valid for a multitude of variations that could be used to disguise the origin of the nuclear material in the device. It is envisioned that this methodology could serve as a preprocessing step to a more computationally intensive and more accurate system in the event of a nuclear terrorist attack.

LaFleur, Adrienne; Charlton, William

2007-09-17T23:59:59.000Z

324

Fast and Thermal Data Testing of LEU, IEU, and HEU Critical Assemblies  

SciTech Connect

The purpose of this paper is to report on data testing of the ENDF/B-VI, release 5, evaluation for LEU, IEU, and HEU benchmarks. In terms of the energy spectrum, there are 10 fast, 3 intermediate, and 21 thermal cases. The characteristics of each benchmark are discussed briefly. The SCALE system (either XSDRN or KENOV.a) with the VITAMIN-B6 (199-group) cross section library were utilized. Hydrogen and U235 from the ENDF/B-VI, release 5, were used in the calculations.

Leal, L.C.; Wright, R.Q.

1999-09-20T23:59:59.000Z

325

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

Science Conference Proceedings (OSTI)

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

326

AN ASSESSMENT OF THE NATIONAL INSTITUTE OF ...  

Science Conference Proceedings (OSTI)

... impact of the hard condensed- matter group in high-critical-temperature ... 7. With the planned transition from highly enriched uranium (HEU) to low ...

2012-09-20T23:59:59.000Z

327

Microsoft Word - 04-ACRONYM LIST.doc  

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

GWSB - Glass Waste Storage Building HATF - High Activity TRU Facility HEU - Highly Enriched Uranium HTF - H-Tank Farm HVAC - Heating, ventilation and air-condition HW - Heavy...

328

Independent Oversight Review of the Idaho National Laboratory...  

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

FHA Fire Hazards Analysis HEU Highly Enriched Uranium HFEF Hot Fuel Examining Facility HSS Office of Health, Safety, and Security HUP High Throughput Uranium Product INL Idaho...

329

PEIS data report: Upgrading the Y-12 Plant for long-term HEU storage  

SciTech Connect

The Department of Energy (DOE) is planning the future of weapons-capable fissile materials owned by the United States (U.S.). Under its Disposition Program, DOE is evaluating its options for: (a) storage of fissile materials needed for specific national programs, and (b) disposal of surplus fissile materials. In accordance with the National Environmental Policy Act (NEPA), DOE is preparing the {open_quotes}Programmatic Environmental Impact Statement (PEIS) for Long-Term Storage and Disposition of Weapons-Usable Fissile Materials{close_quotes} (Disposition PEIS). This paper discusses storage options for highly enriched uranium at the Y-12 plant.

Everitt, D.A.; Johnson, J.P.; Phillips, J.K.; Snider, J.D.

1996-02-01T23:59:59.000Z

330

Pulsed Neutron Measurments With A DT Neutron Generator for an Annular HEU Uranium Metal Casting  

SciTech Connect

Measurements were performed with a single annular, stainless-steel-canned casting of uranium (93.17 wt% 235U) metal ( ~18 kg) to provide data to verify calculational methods for criticality safety. The measurements used a small portable DT generator with an embedded alpha detector to time and directionally tag the neutrons from the generator. The center of the time and directional tagged neutron beam was perpendicular to the axis of the casting. The radiation detectors were 1x1x6 in plastic scintillators encased in 0.635-cm-thick lead shields that were sensitive to neutrons above 1 MeV in energy. The detector lead shields were adjacent to the casting and the target spot of the generator was about 3.8 cm from the casting at the vertical center. The time distribution of the fission induced radiation was measured with respect to the source event by a fast (1GHz) processor. The measurements described in this paper also include time correlation measurements with a time tagged spontaneously fissioning 252Cf neutron source, both on the axis and on the surface of the casting. Measurements with both types of sources are compared. Measurements with the DT generator closely coupled with the HEU provide no more additional information than those with the Cf source closely coupled with the HEU and are complicated by the time and directionally tagged neutrons from the generator scattering between the walls and floor of the measurements room and the casting while still above detection thresholds.

Mihalczo, John T [ORNL; Archer, Daniel E [ORNL; Wright, Michael C [ORNL; Mullens, James Allen [ORNL

2007-09-01T23:59:59.000Z

331

sr0906.docx  

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

workers at H-Canyon recently successfully completed processing the last highly enriched uranium (HEU) components from a Nevada Test Site (NTS) reactor. The NTS reactor,...

332

EIS-0229: Record of Decision (November 2003) | Department of...  

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

for safe and secure storage of weapons-usable fissile materials (plutonium and highly enriched uranium HEU) and a strategy for the disposition of surplus weapons- usable...

333

Gohar, Roglans-Ribas honored for nonproliferation work  

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

mission is to minimize, and to the extent possible eliminate, the use of Highly Enriched Uranium (HEU) in civil applications by converting research and test reactors to the...

334

US-Russia Partnership Reaches Key Milestone in Converting Russian...  

National Nuclear Security Administration (NNSA)

it has monitored the elimination of more than 475 metric tons (MT) of Russian highly enriched uranium (HEU) under a landmark nuclear nonproliferation program, commonly known as...

335

Global Threat  

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

dual application of splitting the atom, U.S. policy towards civilian use of highly enriched uranium (HEU) has historically exhibited contradictory traits. During these early...

336

NNSA Mission Featured on NPR | National Nuclear Security Administratio...  

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

By NNSA Public Affairs NNSA Blog NNSA's successful removal of all remaining highly enriched uranium (HEU) from Ukraine was featured on NPR's "All Things Considered" this past...

337

Audit Report: OAS-L-12-07 | Department of Energy  

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

have recently proven unreliable. In addition, the foreign producers utilize highly enriched uranium (HEU), a practice contrary to the National Nuclear Security Administration's...

338

Nuclear & Radiological Material Removal | National Nuclear Security...  

National Nuclear Security Administration (NNSA)

efforts result in permanent threat reduction. NNSA is returning Russian-origin highly enriched uranium (HEU) fresh and spent fuel from Russian-designed research reactors worldwide...

339

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

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

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

340

Executive Bios: Dr. Jordi Roglans-Ribas - Nuclear Engineering...  

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

mission is to minimize, and to the extent possible eliminate, the use of Highly Enriched Uranium (HEU) in civil applications by converting research and test reactors to the...

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


341

U.S. Department of Energy and NTI Announce Key Nonproliferation...  

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

with the Government of Kazakhstan to move forward with the down-blending of highly enriched uranium (HEU) currently stored at Kazakhstan's Institute of Nuclear Physics. The...

342

EIS-0229: Record of Decision (January 1997) | Department of Energy  

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

for safe and secure storage of weapons-usable fissile materials (plutonium and highly enriched uranium HEU) and a strategy for the disposition of surplus weapons- usable...

343

Record of Decision of the Final Site-Wide Environmental Impact...  

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

4 (No Action-Planning Basis Operations Plus Construct and Operate a Highly Enriched Uranium (HEU) Materials Facility and Special Materials Complex). This alternative...

344

Joint Report Issued by the U.S. Secretary of Energy and the Director...  

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

the aggressive timeline of prioritized repatriation of fresh and spent highly enriched uranium (HEU) fuel and conversion of research reactors in third world countries,"...

345

SR0601  

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

in the process of turning weapons-grade uranium into electricity. Under SRS' Highly Enriched Uranium (HEU) Blend Down program, the preparation and shipping of this material has...

346

Savannah River Operations Office  

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

is estimated the amount of weapons grade material converted at SRS from both highly enriched uranium (HEU) and plutonium could generate enough electricity to power all South...

347

Global Threat Reduction Initiative ? Conversion Program: Reduced...  

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

dual application of splitting the atom, U.S. policy towards civilian use of highly enriched uranium (HEU) has historically exhibited contradictory traits. During these early...

348

DOE/IG-0448 AUDIT REPORT U.S. DEPARTMENT OF ENERGY OFFICE OF...  

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

Materials Protection, Control and Accounting (MPC&A), 2) High Enriched Uranium (HEU), 3 Nuclear Cities Initiative, and 4) Initiative of Proliferation Prevention (IPP). X X X X...

349

NNSA NEWS DRAFT October final edits 19 2009  

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

more than a decade to downblend more than 12 metric tons - enough for approximately 480 nuclear weapons - of excess Russian highly enriched uranium (HEU). NNSA Provides Support...

350

Uranium Mining and Enrichment  

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

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

351

Stable Isotope Enrichment Capabilities at ORNL  

SciTech Connect

The Oak Ridge National Laboratory (ORNL) and the US Department of Energy Nuclear Physics Program have built a high-resolution Electromagnetic Isotope Separator (EMIS) as a prototype for reestablishing a US based enrichment capability for stable isotopes. ORNL has over 60 years of experience providing enriched stable isotopes and related technical services to the international accelerator target community, as well as medical, research, industrial, national security, and other communities. ORNL is investigating the combined use of electromagnetic and gas centrifuge isotope separation technologies to provide research quantities (milligram to several kilograms) of enriched stable isotopes. In preparation for implementing a larger scale production facility, a 10 mA high-resolution EMIS prototype has been built and tested. Initial testing of the device has simultaneously collected greater than 98% enriched samples of all the molybdenum isotopes from natural abundance feedstock.

Egle, Brian [ORNL; Aaron, W Scott [ORNL; Hart, Kevin J [ORNL

2013-01-01T23:59:59.000Z

352

SELECTED STUDIES OF PAST OPERATIONS AT THE ORNL HIGH FLUX ISOTOPE REACTOR  

Science Conference Proceedings (OSTI)

In response to on-going programs at Oak Ridge National Laboratory, two topics related to past operations of the High Flux Isotope Reactor (HFIR) are being reviewed and include determining whether HFIR fuel can be converted from high enriched uranium (HEU) to low enriched uranium (LEU) and determining whether HFIR beryllium reflectors are discharged as transuranic (TRU) waste. The LEU conversion and TRU waste studies are being performed in accordance with the Reduced Enrichment for Research and Test Reactors program and the Integrated Facility Disposition Project, respectively. While assessing data/analysis needs for LEU conversion such as the fuel cycle length and power needed to maintain the current level of reactor performance, a reduction of about 8% (~200 MWD) in the end-of-cycle exposure for HFIR fuel was observed over the lifetime of the reactor (43 years). The SCALE 6.0 computational system was used to evaluate discharged beryllium reflectors and it was discovered if the reflectors are procured according to the current HFIR standard, discharged reflectors would not be TRU waste, but the removable reflector (closest to core) would become TRU waste approximately 40 years after discharge. However, beryllium reflectors have been fabricated with a greater uranium content than that stipulated in the standard and these reflectors would be discharged as TRU waste.

Chandler, David [ORNL; Primm, Trent [ORNL

2010-01-01T23:59:59.000Z

353

Possibility of nuclear pumped laser experiment using low enriched uranium  

SciTech Connect

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

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

2012-06-06T23:59:59.000Z

354

Multiphysics Simulations of the Complex 3D Geometry of the High Flux Isotope Reactor Fuel Elements Using COMSOL  

Science Conference Proceedings (OSTI)

A research and development project is ongoing to convert the currently operating High Flux Isotope Reactor (HFIR) of Oak Ridge National Laboratory (ORNL) from highly-enriched Uranium (HEU U3O8) fuel to low-enriched Uranium (LEU U-10Mo) fuel. Because LEU HFIR-specific testing and experiments will be limited, COMSOL is chosen to provide the needed multiphysics simulation capability to validate against the HEU design data and calculations, and predict the performance of the LEU fuel for design and safety analyses. The focus of this paper is on the unique issues associated with COMSOL modeling of the 3D geometry, meshing, and solution of the HFIR fuel plate and assembled fuel elements. Two parallel paths of 3D model development are underway. The first path follows the traditional route through examination of all flow and heat transfer details using the Low-Reynolds number k-e turbulence model provided by COMSOL v4.2. The second path simplifies the fluid channel modeling by taking advantage of the wealth of knowledge provided by decades of design and safety analyses, data from experiments and tests, and HFIR operation. By simplifying the fluid channel, a significant level of complexity and computer resource requirements are reduced, while also expanding the level and type of analysis that can be performed with COMSOL. Comparison and confirmation of validity of the first (detailed) and second (simplified) 3D modeling paths with each other, and with available data, will enable an expanded level of analysis. The detailed model will be used to analyze hot-spots and other micro fuel behavior events. The simplified model will be used to analyze events such as routine heat-up and expansion of the entire fuel element, and flow blockage. Preliminary, coarse-mesh model results of the detailed individual fuel plate are presented. Examples of the solution for an entire fuel element consisting of multiple individual fuel plates produced by the simplified model are also presented.

Freels, James D [ORNL; Jain, Prashant K [ORNL

2011-01-01T23:59:59.000Z

355

Manuscript under review. Please do not quote without permission. Better Safe than Sorry  

E-Print Network (OSTI)

. The detectors, and the system to use them, aim to find highly enriched uranium (HEU) or plutonium hidden effort to detect highly enriched uranium or plutonium that terrorists could potentially hide

356

A neutronic feasibility study for LEU conversion of the high flux isotope reactor (HFIR).  

SciTech Connect

A neutronic feasibility study was performed to determine the uranium densities that would be required to convert the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) from HEU (93%) to LEU (<20%)fuel. The LEU core that was studied is the same as the current HEU core, except for potential changes in the design of the fuel plates. The study concludes that conversion of HFIR from HEU to LEU fuel would require an advanced fuel with a uranium density of 6-7 gU/cm{sup 3} in the inner fuel element and 9-10 gU/cm{sup 3} in the outer fuel element to match the cycle length of the HEU core. LEU fuel with uranium density up to 4.8 gU/cm{sup 3} is currently qualified for research reactor use. Modifications in fuel grading and burnable poison distribution are needed to produce an acceptable power distribution.

Mo, S. C.

1998-01-14T23:59:59.000Z

357

Critical experiments with 4. 31 wt % /sup 235/U-enriched UO/sub 2/ rods in highly borated water lattices  

Science Conference Proceedings (OSTI)

A series of critical experiments were performed with 4.31 wt % /sup 235/U enriched UO/sub 2/ fuel rods immersed in water containing various concentrations of boron ranging up to 2.55 g/l. The boron was added in the form of boric acid (H/sub 3/BO/sub 3/). Critical experimental data were obtained for two different lattice pitches wherein the water-to-uranium oxide volume ratios were 1.59 and 1.09. The experiments provide benchmarks on heavily borated systems for use in validating calculational techniques employed in analyzing fuel shipping casks and spent fuel storage systems that may utilize boron for criticality control.

Durst, B.M.; Bierman, S.R.; Clayton, E.D.

1982-08-01T23:59:59.000Z

358

Global Laser Enrichment  

E-Print Network (OSTI)

APPLICATION – PUBLIC RESPONSES GE-Hitachi Global Laser Enrichment LLC (GLE) hereby submits the additional information requested in the November 19, 2009 letter. Enclosure 1 of this letter contains the responses the questions. A separate letter has been submitted that contains a non-public version of these responses, which contains Export-Controlled and Security-Related Information. If there are any questions regarding this letter and its contents, please do not hesitate to contact myself, or Julie Olivier of my staff at 910-819-4799 or at Julie.Olivier@ge.com.

Uranium Enrichment Branch; Albert E. Kennedy; Albert E. Kennedy; Tammy Orr (gle

2009-01-01T23:59:59.000Z

359

Global Laser Enrichment  

E-Print Network (OSTI)

PUBLIC VERSION GE-Hitachi Global Laser Enrichment LLC (GLE) hereby submits revision 2 of the GLE License Application. Enclosure 1 contains revised Request for Additional Information responses. Enclosure 2 contains revised chapters 1, 2, 3, 5, 7, and 11 of the GLE License Application. Enclosure 3 contains the revised pubic version of the Decommissioning Funding Plan. Non-Public versions of the revised License Application and the Decommissioning Funding Plan have been prepared and will be submitted under separate enclosure. If there are any questions regarding this letter and its contents, please do not hesitate to contact me at 910-819-4799 or at

Julie Olivier; Brian Smith Chief; Uranium Enrichment Branch; Julie Olivier; Tim Johnson (nrc; Tammy Orr (gle; Lori Butler (geh; Jerry Head (geh; Patricia Campbell (geh; Bob Crate (gle; Ken Givens (gle; Tom Owens (gle

2010-01-01T23:59:59.000Z

360

CONVERSION RATIOS IN SLIGHTLY ENRICHED URANIUM, WATER MODERATED LATTICES  

SciTech Connect

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

Tassan, S.

1963-10-31T23:59:59.000Z

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


361

GROTESQUE: Complex Geometric Arrangement of Unreflected HEU (93.15) Metal Pieces  

SciTech Connect

The GROTESQUE experiments were designed specifically to develop and test neutronics for the GEOM subroutine of the 05R code. Two complex arrangements of various highly enriched uranium metal cylinders, rectangular parallelepipeds, and spheres were arranged in a circular formation on a steel diaphragm. A centerpiece was raised remotely through a hole in the steel diaphragm to achieve criticality. The first arrangement consisted of five major units, each major unit consisting of a stack of smaller uranium pieces. The second arrangement utilized nine major units, again consisting of stacks of smaller uranium pieces. The 9-unit arrangement is the only experiment discussed in this evaluation, since the five stack experiment never achieved criticality. The 9-unit arrangement is shown in Figure 1.1. The experiments were performed at the Oak Ridge Critical Experiments Facility (ORCEF) in June 1964. The 9-unit configuration was later used as part of the development process for early versions of KENO and a model representing a variation of this experiment (Sample Problem 7: GROTESQUE without the Diaphragm) is released with modern versions of SCALE for testing the proper installation of the KENO module. An experimental report for the GROTESQUE experiment has not been published; however there are two publications that describe the experiment (References 1 and 2). A separate reportc discussing the conversion of the 05r model into a KENO model was published; however, the author did not consult with the experimenter for GROTESQUE. This report is considered unreliable (except for dimensions) by the experimenter and should not be used to obtain information pertinent to the GROTESQUE experiment. The Oak Ridge Critical Experiments Facility (ORCEF) Logbook 15r contains the primary documentation from the experimenter for this experiment. The GROTESQUE arrangement of nine major units was evaluated an determined to be an acceptable benchmark experiment.

Mackenzie L. Gorham; John D. Bess

2011-09-01T23:59:59.000Z

362

SR-08-03 _LEU Shipments_.doc  

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

up the Site. As part of an interagency agreement with TVA, DOE down blended highly enriched uranium (HEU) and provided a low-enriched uranium (LEU) solution to commercial fuel...

363

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

SciTech Connect

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

Tripard, G.E.

1993-06-01T23:59:59.000Z

364

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

SciTech Connect

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

Tripard, G.E.

1994-12-31T23:59:59.000Z

365

EIS-0240: Record of Decision | Department of Energy  

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

Department of Energy (DOE) has decided to implement a program to make surplus highly enriched uranium (HEU) non-weapons-usable by blending it down to low-enriched uranium (LEU),...

366

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

E-Print Network (OSTI)

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

Romano, Paul K. (Paul Kollath)

2009-01-01T23:59:59.000Z

367

Expanding and optimizing fuel management and data analysis capabilities of MCODE-FM in support of MIT research reactor (MITR-II) LEU conversion  

E-Print Network (OSTI)

Studies are underway in support of the MIT research reactor (MITR-II) conversion from high enriched Uranium (HEU) to low enriched Uranium (LEU), as required by recent non-proliferation policy. With the same core configuration ...

Horelik, Nicholas E. (Nicholas Edward)

2012-01-01T23:59:59.000Z

368

Development of a core design optimization tool and analysis in support of the planned LEU conversion of the MIT Research Reactor (MITR-II)  

E-Print Network (OSTI)

The MIT Research Reactor (MITR-II) is currently undergoing analysis for the planned conversion from high enriched uranium (HEU) to low enriched uranium (LEU), as part of a global effort to minimize the availability of ...

Connaway, Heather M. (Heather Moira)

2012-01-01T23:59:59.000Z

369

Overview: A Legacy of Uranium Enrichment  

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

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

370

Detection of uranium enrichment activities using environmental monitoring techniques  

SciTech Connect

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

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

1993-03-30T23:59:59.000Z

371

Assuaging Nuclear Energy Risks: The Angarsk International Uranium Enrichment Center  

Science Conference Proceedings (OSTI)

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

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

2011-06-28T23:59:59.000Z

372

Advanced Neutron Source enrichment study  

SciTech Connect

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

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

1994-12-31T23:59:59.000Z

373

HEU Holdup Measurements in the 321-M Draw Bench, Straightener, and Fluoroscope Components  

SciTech Connect

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.

Dewberry, R.A.

2001-07-10T23:59:59.000Z

374

Zeus: Fast-spectrum critical assemblies with an iron-HEU core surrounded by a copper reflector  

SciTech Connect

Experiments to investigate critical systems of iron moderated highly enriched uranium in the intermediate-energy range were attempted. However, due to size limitations, the systems fell into the fast-energy range. Two critical configurations were established with a uranium mass of {approx} 198 kg and a Fe/{sup 235}U Ratio of {approx}15. Experimental uncertainties were systematically evaluated to estimate their effect on multiplication. The combined uncertainty for these experiments is estimated to be {+-}0.0024 {Delta}{sub eff}. Consequently, both Zeus iron configurations are judged to be acceptable for use as criticality-safety benchmark experiments. (authors)

Hayes, D. K.; Sanchez, R. G.; Kahler, A. C. [Los Alamos National Laboratory, PO Box 1663, Los Alamos, NM 87545 (United States)

2006-07-01T23:59:59.000Z

375

Isotope Enrichment | ornl.gov  

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

Modern electromagnetic isotope separator developed and being scaled-up to replace the Manhattan Project-era Calutrons used for stable isotope enrichment. Since 1945, ORNL has...

376

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

SciTech Connect

Remote handling equipment is required to achieve the processing of highly radioactive, post reactor, fuel for the melt-dilute process, which will convert high enrichment uranium fuel elements into lower enrichment forms for subsequent disposal. The melt-dilute process combines highly radioactive enriched uranium fuel elements with deleted uranium and aluminum for inductive melting and inductive stirring steps that produce a stable aluminum/uranium ingot of low enrichment.

Heckendorn, F.M.

2001-01-03T23:59:59.000Z

377

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

Science Conference Proceedings (OSTI)

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 the methodology, non-destructive assay (NDA) measurements, and results of the uranium holdup on the two furnaces.

Salaymeh, S.R.

2002-04-30T23:59:59.000Z

378

Low enrichment fuel conversion for Iowa State University  

SciTech Connect

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

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

1990-08-01T23:59:59.000Z

379

Microsoft Word - SRS-WD-2010-001_R0_Final_9-30-10.doc  

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

HA Hazard Analysis HAW High Activity Waste HDPE High Density Polyethylene HEU Highly Enriched Uranium HM H-Modified hr hour(s) HRR Highly Radioactive Radionuclide HTF H-Tank Farm...

380

Uranium Enrichment Decontamination and Decommissioning Fund's...  

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

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

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


381

Enrichment Assay Methods for a UF6 Cylinder Verification Station  

Science Conference Proceedings (OSTI)

International Atomic Energy Agency (IAEA) inspectors currently perform periodic inspections at uranium enrichment plants to verify UF6 cylinder enrichment declarations. Measurements are typically performed with handheld high-resolution sensors on a sampling of cylinders taken to be representative of the facility’s entire cylinder inventory. These enrichment assay methods interrogate only a small fraction of the total cylinder volume, and are time-consuming and expensive to execute for inspectors. Pacific Northwest National Laboratory (PNNL) is developing an unattended measurement system capable of automated enrichment measurements over the full volume of Type 30B and Type 48 cylinders. This Integrated Cylinder Verification System (ICVS) could be located at key measurement points to positively identify each 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 focus of this paper is the development of nondestructive assay (NDA) methods that combine “traditional” enrichment signatures (e.g. 185-keV emission from U-235) and more-penetrating “non-traditional” signatures (e.g. high-energy neutron-induced gamma rays spawned primarily from U-234 alpha emission) collected by medium-resolution gamma-ray spectrometers (i.e. sodium iodide or lanthanum bromide). The potential of these NDA methods for the automated assay of feed, tail and product cylinders is explored through MCNP modeling and with field measurements on a cylinder population ranging from 0.2% to 5% in U-235 enrichment.

Smith, Leon E.; Jordan, David V.; Misner, Alex C.; Mace, Emily K.; Orton, Christopher R.

2010-11-30T23:59:59.000Z

382

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

E-Print Network (OSTI)

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

unknown authors

2003-01-01T23:59:59.000Z

383

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

E-Print Network (OSTI)

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

American Society for Testing and Materials. Philadelphia

2000-01-01T23:59:59.000Z

384

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

SciTech Connect

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

Travelli, A.

1983-01-01T23:59:59.000Z

385

Enrichment of light hydrocarbon mixture  

Science Conference Proceedings (OSTI)

Light hydrocarbon enrichment is accomplished using a vertically oriented distillation column having a plurality of vertically oriented, nonselective micro/mesoporous hollow fibers. Vapor having, for example, both propylene and propane is sent upward through the distillation column in between the hollow fibers. Vapor exits neat the top of the column and is condensed to form a liquid phase that is directed back downward through the lumen of the hollow fibers. As vapor continues to ascend and liquid continues to countercurrently descend, the liquid at the bottom of the column becomes enriched in a higher boiling point, light hydrocarbon (propane, for example) and the vapor at the top becomes enriched in a lower boiling point light hydrocarbon (propylene, for example). The hollow fiber becomes wetted with liquid during the process.

Yang, Dali (Los Alamos, NM); Devlin, David (Santa Fe, NM); Barbero, Robert S. (Santa Cruz, NM); Carrera, Martin E. (Naperville, IL); Colling, Craig W. (Warrenville, IL)

2011-11-29T23:59:59.000Z

386

Enrichment of light hydrocarbon mixture  

DOE Patents (OSTI)

Light hydrocarbon enrichment is accomplished using a vertically oriented distillation column having a plurality of vertically oriented, nonselective micro/mesoporous hollow fibers. Vapor having, for example, both propylene and propane is sent upward through the distillation column in between the hollow fibers. Vapor exits neat the top of the column and is condensed to form a liquid phase that is directed back downward through the lumen of the hollow fibers. As vapor continues to ascend and liquid continues to countercurrently descend, the liquid at the bottom of the column becomes enriched in a higher boiling point, light hydrocarbon (propane, for example) and the vapor at the top becomes enriched in a lower boiling point light hydrocarbon (propylene, for example). The hollow fiber becomes wetted with liquid during the process.

Yang; Dali (Los Alamos, NM); Devlin, David (Santa Fe, NM); Barbero, Robert S. (Santa Cruz, NM); Carrera, Martin E. (Naperville, IL); Colling, Craig W. (Warrenville, IL)

2010-08-10T23:59:59.000Z

387

AVLIS enrichment of medical isotopes  

SciTech Connect

Under the Sponsorship of the United states Enrichment Corporation (USEC), we are currently investigating the large scale separation of several isotopes of medical interest using atomic vapor isotope separation (AVLIS). This work includes analysis and experiments in the enrichment of thallium 203 as a precursor to the production of thallium 201 used in cardiac imaging following heart attacks, on the stripping of strontium 84 from natural strontium as precursor to the production of strontium 89, and on the stripping of lead 210 from lead used in integrated circuits to reduce the number of alpha particle induced logic errors.

Haynam, C.A.; Scheibner, K.F.; Stern, R.C.; Worden, E.F. [Lawrence Livermore National Laboratory, CA (United States)

1996-12-31T23:59:59.000Z

388

Uranium enrichment in the United States  

SciTech Connect

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

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

1975-01-01T23:59:59.000Z

389

Disposition of Surplus Highly Enriched Uranium  

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

. . ------- .--- --. ---- DOE/EIS-0240 I United States Department of Energy I For Further Information Contact: U.S. Department of Energy Otice of Fissile Materials Disposition, 1000 Independence Ave., SW, Washington, D.C. 20585 1 I ---- I I . I I I I This report has been reproduced directly from the best available copy. Available to DOE and DOE contractors from the Office of Scientific and Technical Information, P.O. Box 62, Oak Ridge, TN 37831; telephone (423) 576-8401 for prices. Available to the public from the National Technical Information Service, 5285 Port Royal Road, Springfield, VA 22161. Copies of this document are available (while supplies last) upon written request to: I Office of Fissile Materials Disposition, MD-4 Forrestal Building United States Department of Energy 1000 Independence Avenue, SW Washington, DC 20585 , @ Printed with soy ink on recycled paper. -_. - COVERS~ET

390

Tritium production using highly enriched fuel  

SciTech Connect

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

Miller, R.L.

1967-12-18T23:59:59.000Z

391

Highly Enriched Uranium Transparency Program | National Nuclear...  

National Nuclear Security Administration (NNSA)

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

392

Disposition of Surplus Highly Enriched Uranium  

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

EIS-0240-S EIS-0240-S For Further Information Contact: U.S. Departmel>t of Energy Office of Fissile Materials Disposition, 1000 Independence Ave., SW, Washington, D.C. 20585 . This report has been reproduced directly from the best available copy. Available to DOE and DOE contractors from the Office of Scientific and Technical Information, P.O. Box 62, Oak Ridge, TN 37831; telephone (423) 576-8401 for prices, Available to the public from the National Technical Information Service, 5285 Port Royal Road, Springfield, VA 22161. Copies of this document are available (while supplies last) upon written request to: Office of Fissile Materials Disposition, MD-4 Forrestal Building United States Department of Energy 1000 Independence Avenue, SW Washington, DC 20585 @ Printed with soy ink on recycled paper. .__- -. @ .: Depafimmt of Energy . i i~t " Wastin@on, DC 20585 June 1996 Dear hterested

393

Progress report on the conversion of the Purdue University reactor, PUR-1, from HEU to LEU  

Science Conference Proceedings (OSTI)

The effort for the conversion of the Purdue University Research Reactor, PUR-1, began in August, 2005, and will be completed in late 2007. Initial low-enriched uranium (LEU) assemblies will be inserted into the core in September, and the final core load is expected to be completed by October 2007. This paper summarizes the work performed to date, and the expectations for the work remaining to complete the project. The PUR-1 conversion has been a collaborative effort with Purdue, Idaho National Laboratory, Argonne National Laboratory participating under the auspices of the U.S. Department of Energy-Global Threat Reduction. Construction of PUR-1 began in 1961, and it was initially licensed by the U.S. Nuclear Regulatory Commission in August of 1962 for operation at 1kW. The primary missions of the reactor are training, education and research. Each graduate of the School of Nuclear Engineering at Purdue University will have operated the reactor in two or more separate experiments as part of their curriculum. Students were used extensively throughout all phases of this conversion project, providing additional learning opportunities to complete their education experience. (author)

Jenkins, J.H.; Merritt, E.C.; Revis, B. [School of Nuclear Engineering, Purdue University, 400 Central Dr., West Lafayette, IN 47907 (United States)

2008-07-15T23:59:59.000Z

394

A Robust Infrastructure Design for Gas Centrifuge Enrichment Plant Unattended Online Enrichment Monitoring  

Science Conference Proceedings (OSTI)

An online enrichment monitor (OLEM) is being developed to continuously measure the relative isotopic composition of UF6 in the unit header pipes of a gas centrifuge enrichment plant (GCEP). From a safeguards perspective, OLEM will provide early detection of a facility being misused for production of highly enriched uranium. OLEM may also reduce the number of samples collected for destructive assay and if coupled with load cell monitoring can provide isotope mass balance verification. The OLEM design includes power and network connections for continuous monitoring of the UF6 enrichment and state of health of the instrument. Monitoring the enrichment on all header pipes at a typical GCEP could require OLEM detectors on each of the product, tails, and feed header pipes. If there are eight process units, up to 24 detectors may be required at a modern GCEP. Distant locations, harsh industrial environments, and safeguards continuity of knowledge requirements all place certain demands on the network robustness and power reliability. This paper describes the infrastructure and architecture of an OLEM system based on OLEM collection nodes on the unit header pipes and power and network support nodes for groupings of the collection nodes. A redundant, self-healing communications network, distributed backup power, and a secure communications methodology. Two candidate technologies being considered for secure communications are the Object Linking and Embedding for Process Control Unified Architecture cross-platform, service-oriented architecture model for process control communications and the emerging IAEA Real-time And INtegrated STream-Oriented Remote Monitoring (RAINSTORM) framework to provide the common secure communication infrastructure for remote, unattended monitoring systems. The proposed infrastructure design offers modular, commercial components, plug-and-play extensibility for GCEP deployments, and is intended to meet the guidelines and requirements for unattended and remotely monitored safeguards systems.

Younkin, James R [ORNL; Rowe, Nathan C [ORNL; Garner, James R [ORNL

2012-01-01T23:59:59.000Z

395

Low-Enrichment Fuel Development Program  

SciTech Connect

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

Stahl, D.

1978-01-01T23:59:59.000Z

396

The complexity of enriched µ-calculi  

Science Conference Proceedings (OSTI)

The fully enriched ?-calculus is the extension of the propositional ?-calculus with inverse programs, graded modalities, and nominals. While satisfiability in several expressive fragments of the fully enriched ?-calculus ...

Piero A. Bonatti; Carsten Lutz; Aniello Murano; Moshe Y. Vardi

2006-07-01T23:59:59.000Z

397

--No Title--  

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

DOE United States Department of Energy GSMS UF 6 Gas Mass Spectrometry HEU Highly-enriched uranium ( 20 weight percent 235 U) ICP-MS Inductively Coupled Plasma Mass...

398

Microsoft Word - 09-Chap 4 Hazard Specific.doc  

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

See Figure 4.1, EM- owned Nuclear Materials F, H and K Areas Uranium (U), Highly Enriched Uranium (HEU) and Depleted Uranium (DU) Uranium nuclear materials are being...

399

[Cover page, Margins: Left 1 in  

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

respectively. The fuel plates, which are involute in shape, are made of highly enriched uranium (HEU) in a mixture of U 3 O 8 -Al sandwiched between two sheets of Al-6061....

400

Belgium Nuclear Security Summit: Fact Sheet | National Nuclear...  

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

announced it will work jointly with the United States to eliminate its excess highly enriched uranium (HEU) and plutonium. SCK-CEN has been working closely for the past year with...

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


401

THE THEORY OF URANIUM ENRICHMENT BY THE GAS CENTRIFUGE  

E-Print Network (OSTI)

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

Olander, Donald R.

2013-01-01T23:59:59.000Z

402

THE THEORY OF URANIUM ENRICHMENT BY THE GAS CENTRIFUGE  

E-Print Network (OSTI)

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

Olander, Donald R.

2013-01-01T23:59:59.000Z

403

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

SciTech Connect

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

1993-07-01T23:59:59.000Z

404

133Ba as a gamma-ray surrogate source for 1kg HEU and 10g 239Pu and 252Cf as a Neutron Surrogate for Pu  

SciTech Connect

Monte Carlo was performed for the purpose of relating gamma-ray signal strength from 1kg of HEU and 10g of {sup 239}Pu (as described in the ASTM standards) to the radiation emitted from an amount of {sup 133}Ba. A determination was made on the amount of {sup 133}Ba that could act as a surrogate for the specified amounts of HEU and Pu. {sup 133}Ba is not the ideal source to use as a surrogate for HEU because of its higher energies. {sup 133}Ba was chosen as the surrogate since it has a half-life of 10.54 years, rather then the more ideal surrogate of {sup 57}Co which has a half-life of 271 days. A similar Monte Carlo was performed for the purpose of relating neutron signal strength from 200g of Pu (as described in the ASTM standards) to the radiation emitted from an amount of shielded {sup 252}Cf. A determination was made on the amount of {sup 252}Cf necessary to act as a surrogate for the 200g of Pu. An ASTM standard source is a metallic sphere, cube, or right cylinder of SNM having maximum self-attenuation of its emitted radiation. For plutonium, the source should be at least 93% {sup 239}Pu, less than 6.5% {sup 240}Pu, and less than 0.5% impurities. A cadmium filter of at least 0.08cm thick should be used to reduce the impact of {sup 241}Am. For uranium, the source should contain at least 95% {sup 235}U and less than 0.25% impurities. For neutron detector testing, the neutron source shall be placed in a lead shielding container that reduces the gamma radiation from the source to 1% of its unshielded value.

Pohl, B A; Archer, D E

2004-03-11T23:59:59.000Z

405

RERTR program reduces use of enriched uranium in research reactors  

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

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

406

Enrichment Zoning Options for the Small Nuclear Rocket Engine (SNRE)  

SciTech Connect

Advancement of U.S. scientific, security, and economic interests through a robust space exploration program requires high performance propulsion systems to support a variety of robotic and crewed missions beyond low Earth orbit. In NASA’s recent Mars Design Reference Architecture (DRA) 5.0 study (NASA-SP-2009-566, July 2009), nuclear thermal propulsion (NTP) was again selected over chemical propulsion as the preferred in-space transportation system option because of its high thrust and high specific impulse (-900 s) capability, increased tolerance to payload mass growth and architecture changes, and lower total initial mass in low Earth orbit. An extensive nuclear thermal rocket technology development effort was conducted from 1955-1973 under the Rover/NERVA Program. The Small Nuclear Rocket Engine (SNRE) was the last engine design studied by the Los Alamos National Laboratory during the program. At the time, this engine was a state-of-the-art design incorporating lessons learned from the very successful technology development program. Past activities at the NASA Glenn Research Center have included development of highly detailed MCNP Monte Carlo transport models of the SNRE and other small engine designs. Preliminary core configurations typically employ fuel elements with fixed fuel composition and fissile material enrichment. Uniform fuel loadings result in undesirable radial power and temperature profiles in the engines. Engine performance can be improved by some combination of propellant flow control at the fuel element level and by varying the fuel composition. Enrichment zoning at the fuel element level with lower enrichments in the higher power elements at the core center and on the core periphery is particularly effective. Power flattening by enrichment zoning typically results in more uniform propellant exit temperatures and improved engine performance. For the SNRE, element enrichment zoning provided very flat radial power profiles with 551 of the 564 fuel elements within 1% of the average element power. Results for this and alternate enrichment zoning options for the SNRE are compared.

Bruce G. Schnitzler; Stanley K. Borowski

2010-07-01T23:59:59.000Z

407

Novel Membranes and Processes for Oxygen Enrichment  

SciTech Connect

The overall goal of this project is to develop a membrane process that produces air containing 25-35% oxygen, at a cost of $25-40/ton of equivalent pure oxygen (EPO2). Oxygen-enriched air at such a low cost will allow existing air-fueled furnaces to be converted economically to oxygen-enriched furnaces, which in turn will improve the economic and energy efficiency of combustion processes significantly, and reduce the cost of CO{sub 2} capture and sequestration from flue gases throughout the U.S. manufacturing industries. During the 12-month Concept Definition project: We identified a series of perfluoropolymers (PFPs) with promising oxygen/nitrogen separation properties, which were successfully made into thin film composite membranes. The membranes showed oxygen permeance as high as 1,200 gpu and oxygen/nitrogen selectivity of 3.0, and the permeance and selectivity were stable over the time period tested (60 days). We successfully scaled up the production of high-flux PFP-based membranes, using MTR's commercial coaters. Two bench-scale spiral-wound modules with countercurrent designs were made and parametric tests were performed to understand the effect of feed flow rate and pressure, permeate pressure and sweep flow rate on the membrane module separation properties. At various operating conditions that modeled potential industrial operating conditions, the module separation properties were similar to the pure-gas separation properties in the membrane stamps. We also identified and synthesized new polymers [including polymers of intrinsic microporosity (PIMs) and polyimides] with higher oxygen/nitrogen selectivity (3.5-5.0) than the PFPs, and made these polymers into thin film composite membranes. However, these membranes were susceptible to severe aging; pure-gas permeance decreased nearly six-fold within two weeks, making them impractical for industrial applications of oxygen enrichment. We tested the effect of oxygen-enriched air on NO{sub x} emissions using a Bloom baffle burner at GTI. The results are positive and confirm that oxygen-enriched combustion can be carried out without producing higher levels of NOx than normal air firing, if lancing of combustion air is used and the excess air levels are controlled. A simple economic study shows that the membrane processes can produce O{sub 2} at less than $40/ton EPO{sub 2} and an energy cost of 1.1-1.5 MMBtu/ton EPO{sub 2}, which are very favorable compared with conventional technologies such as cryogenics and vacuum pressure swing adsorption processes. The benefits of integrated membrane processes/combustion process trains have been evaluated, and show good savings in process costs and energy consumption, as well as reduced CO{sub 2} emissions. For example, if air containing 30% oxygen is used in natural gas furnaces, the net natural gas savings are an estimated 18% at a burner temperature of 2,500 F, and 32% at a burner temperature of 3,000 F. With a 20% market penetration of membrane-based oxygen-enriched combustion in all combustion processes by 2020, the energy savings would be 414-736 TBtu/y in the U.S. The comparable net cost savings are estimated at $1.2-2.1 billion per year by 2020, calculated as the value of fuel savings subtracted from the cost of oxygen production. The fuel savings of 18%-32% by the membrane/oxygen-enriched combustion corresponds to an 18%-32% reduction in CO{sub 2} emissions, or 23-40 MM ton/y less CO{sub 2} from natural gas-fired furnaces by 2020. In summary, results from this project (Concept Definition phase) are highly promising and clearly demonstrate that membrane processes can produce oxygen-enriched air in a low cost manner that will lower operating costs and energy consumption in industrial combustion processes. Future work will focus on proof-of-concept bench-scale demonstration in the laboratory.

Lin, Haiqing

2011-11-15T23:59:59.000Z

408

Automated UF6 Cylinder Enrichment Assay: Status of the Hybrid Enrichment Verification Array (HEVA) Project: POTAS Phase II  

SciTech Connect

Pacific Northwest National Laboratory (PNNL) intends to automate the UF6 cylinder nondestructive assay (NDA) verification currently performed by the International Atomic Energy Agency (IAEA) at enrichment plants. PNNL is proposing the installation of a portal monitor at a key measurement point to positively identify each cylinder, measure its mass and enrichment, store the data along with operator inputs in a secure database, and maintain continuity of knowledge on measured cylinders until inspector arrival. This report summarizes the status of the research and development of an enrichment assay methodology supporting the cylinder verification concept. The enrichment assay approach exploits a hybrid of two passively-detected ionizing-radiation signatures: the traditional enrichment meter signature (186-keV photon peak area) and a non-traditional signature, manifested in the high-energy (3 to 8 MeV) gamma-ray continuum, generated by neutron emission from UF6. PNNL has designed, fabricated, and field-tested several prototype assay sensor packages in an effort to demonstrate proof-of-principle for the hybrid assay approach, quantify the expected assay precision for various categories of cylinder contents, and assess the potential for unsupervised deployment of the technology in a portal-monitor form factor. We refer to recent sensor-package prototypes as the Hybrid Enrichment Verification Array (HEVA). The report provides an overview of the assay signatures and summarizes the results of several HEVA field measurement campaigns on populations of Type 30B UF6 cylinders containing low-enriched uranium (LEU), natural uranium (NU), and depleted uranium (DU). Approaches to performance optimization of the assay technique via radiation transport modeling are briefly described, as are spectroscopic and data-analysis algorithms.

Jordan, David V.; Orton, Christopher R.; Mace, Emily K.; McDonald, Benjamin S.; Kulisek, Jonathan A.; Smith, Leon E.

2012-06-01T23:59:59.000Z

409

Reduced enrichment for research and test reactors: Proceedings  

SciTech Connect

The international effort to develop new research reactor fuel materials and designs based on the use of low-enriched uranium, instead of highly-enriched uranium, has made much progress during the eight years since its inception. To foster direct communication and exchange of ideas among the specialist in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the ninth of a series which began in 1978. All previous meetings of this series are listed on the facing page. The focus of this meeting was on the LEU fuel demonstration which was in progress at the Oak Ridge Research (ORR) reactor, not far from where the meeting was held. The visit to the ORR, where a silicide LEU fuel with 4.8 g A/cm/sup 3/ was by then in routine use, illustrated how far work has progressed.

1988-05-01T23:59:59.000Z

410

Reactor and Material Supply | Y-12 National Security Complex  

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

Reactor and Reactor and Material Supply Reactor and Material Supply Y-12 has processed highly enriched uranium for more than 60 years in support of the nation's defense. The end of the Cold War and ensuing strategic arms control treaties have resulted in an excess of HEU materials. In 1994, approximately 174 metric tons of weapons-usable HEU was declared surplus to defense needs. The HEU disposition program was established to make the surplus HEU unsuitable for use in weapons by blending it down to low-enriched uranium and to recover the economic value of the materials to the extent practical. In 2005, the Secretary of Energy announced that an additional 200 metric tons of HEU would be removed from further use as fissile material in U.S. nuclear weapons. Approximately 20 metric tons of this material will

411

Prompt Neutron Lifetime for the NBSR Reactor  

SciTech Connect

In preparation for the proposed conversion of the National Institute of Standards and Technology (NIST) research reactor (NBSR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel, certain point kinetics parameters must be calculated. We report here values of the prompt neutron lifetime that have been calculated using three independent methods. All three sets of calculations demonstrate that the prompt neutron lifetime is shorter for the LEU fuel when compared to the HEU fuel and longer for the equilibrium end-of-cycle (EOC) condition when compared to the equilibrium startup (SU) condition for both the HEU and LEU fuels.

Hanson, A.L.; Diamond, D.

2012-06-24T23:59:59.000Z

412

Health and safety considerations for U. S. monitors in the Russian transparency program.  

SciTech Connect

In 1993 the US and the Russian Federation signed an agreement allowing the US to purchase highly enriched uranium (HEU) from Russia over a 20-year period. This Highly Enriched Uranium Purchase Agreement permits the purchase of 500 metric tons of HEU from dismantled Russian nuclear weapons in the form of low-enriched uranium (LEU) for use as power reactor fuel in the US. Under the HEU Agreement, the US and Russia are cooperating in a ''Transparency Program'' to ensure that arms control and nonproliferation objectives are being met. The Transparency Program measures, which are a departure from traditional, intrusive measures of verification, include sending individuals from the US to Russia to monitor the processing of the HEU.

Boggs, C. J.

1998-10-22T23:59:59.000Z

413

Table 17. Purchases of enrichment services by owners and operators ...  

U.S. Energy Information Administration (EIA)

Next Release Date: May 2014 Enrichment Service Contract Type: U.S. Enrichment Foreign Enrichment: Total Spot : 0 521 : 521 Long-Term : 3,261 11,808 : 15,069

414

Enriched Analyses with Assimilation of SALLJEX Data  

Science Conference Proceedings (OSTI)

This paper aims at generating a set of enriched analyses by assimilating the data collected during the South American Low-Level Jet Experiment (SALLJEX) in southeastern South America during the summer season 2002/03. The analyses are generated ...

Yanina García Skabar; Matilde Nicolini

2009-12-01T23:59:59.000Z

415

U. S. forms uranium enrichment corporation  

SciTech Connect

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

Seltzer, R.

1993-07-12T23:59:59.000Z

416

SURVEY OF LOW ENRICHMENT MOLTEN-SALT REACTORS  

SciTech Connect

A rough survey of the nuclear charactenistics of graphite-moderated molten-salt reactors utilizing an initial complement of low enrichment uranium fuel has been made. Reactors can be constructed with initial enrichinents as low as 1.25% U-235; initial conversion ratios of as high as 0.8 can be obtained with enrichinent of less than 2%. Highly enriched uraninm would be added as make-up fuel, and such reactors could probably be operated for bunnups as high as 60,000 Mwd/ton before buildup of fission preducts wpuld make replacement of the fuel desirable. A typical circulating fuel reactor of this class might contain an initial inventory of 3600 tons of 1.8% enriched uranium, operated at 640 Mw (thermal), and generate a net of 260 Mw (electrical). The total fuel cycle cost would be approximately 1.3 mills/kwhr, of which 1.0 mill is bunnup of enniched U- 235. (auth)

MacPherson, H.G.

1958-10-17T23:59:59.000Z