National Library of Energy BETA

Sample records for tritium extraction facility

  1. TRITIUM EXTRACTION FACILITY ALARA

    SciTech Connect (OSTI)

    Joye, BROTHERTON

    2005-04-19

    The primary mission of the Tritium Extraction Facility (TEF) is to extract tritium from tritium producing burnable absorber rods (TPBARs) that have been irradiated in a commercial light water reactor and to deliver tritium-containing gas to the Savannah River Site Facility 233-H. The tritium extraction segment provides the capability to deliver three (3) kilograms per year to the nation's nuclear weapons stockpile. The TEF includes processes, equipment and facilities capable of production-scale extraction of tritium while minimizing personnel radiation exposure, environmental releases, and waste generation.

  2. Calibrations of a tritium extraction facility

    SciTech Connect (OSTI)

    Bretscher, M.M.; Oliver, B.M.; Farrar, H. IV

    1983-01-01

    A tritium extraction facility has been built for the purpose of measuring the absolute tritium concentration in neutron-irradiated lithium metal samples. Two independent calibration procedures have been used to determine what fraction, if any, of tritium is lost during the extraction process. The first procedure compares independently measured /sup 4/He and /sup 3/H concentrations from the /sup 6/Li(n,..cap alpha..)T reaction. The second procedure compared measured /sup 6/Li(n,..cap alpha..)T//sup 197/Au (n,..gamma..)/sup 198/Au thermal neutron reaction rate ratios with those obtained from Monte Carlo calculations using well-known cross sections. Both calibration methods show that within experimental errors (approx. 1.5%) no tritium is lost during the extraction process.

  3. Commercial Light Water Reactor Tritium Extraction Facility

    SciTech Connect (OSTI)

    McHood, M D

    2000-10-12

    A geotechnical investigation program has been completed for the Commercial Light Water Reactor - Tritium Extraction Facility (CLWR-TEF) at the Savannah River Site (SRS). The program consisted of reviewing previous geotechnical and geologic data and reports, performing subsurface field exploration, field and laboratory testing, and geologic and engineering analyses. The purpose of this investigation was to characterize the subsurface conditions for the CLWR-TEF in terms of subsurface stratigraphy and engineering properties for design and to perform selected engineering analyses. The objectives of the evaluation were to establish site-specific geologic conditions, obtain representative engineering properties of the subsurface and potential fill materials, evaluate the lateral and vertical extent of any soft zones encountered, and perform engineering analyses for slope stability, bearing capacity and settlement, and liquefaction potential. In addition, provide general recommendations for construction and earthwork.

  4. Structural acceptance criteria Remote Handling Building Tritium Extraction Facility

    SciTech Connect (OSTI)

    Mertz, G.

    1999-12-16

    This structural acceptance criteria contains the requirements for the structural analysis and design of the Remote Handling Building (RHB) in the Tritium Extraction Facility (TEF). The purpose of this acceptance criteria is to identify the specific criteria and methods that will ensure a structurally robust building that will safely perform its intended function and comply with the applicable Department of Energy (DOE) structural requirements.

  5. Commercial Light Water Reactor Tritium Extraction Facility Geotechnical Summary Report

    SciTech Connect (OSTI)

    Lewis, M.R.

    2000-01-11

    A geotechnical investigation program has been completed for the Circulating Light Water Reactor - Tritium Extraction Facility (CLWR-TEF) at the Savannah River Site (SRS). The program consisted of reviewing previous geotechnical and geologic data and reports, performing subsurface field exploration, field and laboratory testing and geologic and engineering analyses. The purpose of this investigation was to characterize the subsurface conditions for the CLWR-TEF in terms of subsurface stratigraphy and engineering properties for design and to perform selected engineering analyses. The objectives of the evaluation were to establish site-specific geologic conditions, obtain representative engineering properties of the subsurface and potential fill materials, evaluate the lateral and vertical extent of any soft zones encountered, and perform engineering analyses for slope stability, bearing capacity and settlement, and liquefaction potential. In addition, provide general recommendations for construction and earthwork.

  6. STAR Facility Tritium Accountancy

    SciTech Connect (OSTI)

    R. J. Pawelko; J. P. Sharpe; B. J. Denny

    2007-09-01

    The Safety and Tritium Applied Research (STAR) facility has been established to provide a laboratory infrastructure for the fusion community to study tritium science associated with the development of safe fusion energy and other technologies. STAR is a radiological facility with an administrative total tritium inventory limit of 1.5g (14,429 Ci) [1]. Research studies with moderate tritium quantities and various radionuclides are performed in STAR. Successful operation of the STAR facility requires the ability to receive, inventory, store, dispense tritium to experiments, and to dispose of tritiated waste while accurately monitoring the tritium inventory in the facility. This paper describes tritium accountancy in the STAR facility. A primary accountancy instrument is the tritium Storage and Assay System (SAS): a system designed to receive, assay, store, and dispense tritium to experiments. Presented are the methods used to calibrate and operate the SAS. Accountancy processes utilizing the Tritium Cleanup System (TCS), and the Stack Tritium Monitoring System (STMS) are also discussed. Also presented are the equations used to quantify the amount of tritium being received into the facility, transferred to experiments, and removed from the facility. Finally, the STAR tritium accountability database is discussed.

  7. STAR facility tritium accountancy

    SciTech Connect (OSTI)

    Pawelko, R. J.; Sharpe, J. P.; Denny, B. J.

    2008-07-15

    The Safety and Tritium Applied Research (STAR) facility has been established to provide a laboratory infrastructure for the fusion community to study tritium science associated with the development of safe fusion energy and other technologies. STAR is a radiological facility with an administrative total tritium inventory limit of 1.5 g (14,429 Ci) [1]. Research studies with moderate tritium quantities and various radionuclides are performed in STAR. Successful operation of the STAR facility requires the ability to receive, inventory, store, dispense tritium to experiments, and to dispose of tritiated waste while accurately monitoring the tritium inventory in the facility. This paper describes tritium accountancy in the STAR facility. A primary accountancy instrument is the tritium Storage and Assay System (SAS): a system designed to receive, assay, store, and dispense tritium to experiments. Presented are the methods used to calibrate and operate the SAS. Accountancy processes utilizing the Tritium Cleanup System (TCS), and the Stack Tritium Monitoring System (STMS) are also discussed. Also presented are the equations used to quantify the amount of tritium being received into the facility, transferred to experiments, and removed from the facility. Finally, the STAR tritium accountability database is discussed. (authors)

  8. Independent Oversight Review, Savannah River Site Tritium Facilities...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    June 2012 Independent Oversight Review, Savannah River Site Tritium Facilities - June 2012 June 2012 Review of the Savannah River Site Tritium Facilities Implementation...

  9. Tritium emission reduction at Darlington tritium removal facility using a Bubbler System

    SciTech Connect (OSTI)

    Kalyanam, K.; Leilabadi, A.; El-Behairy, O.; Williams, G. I. D.; Vogt, H. K. [Ontario Power Generation, Darlington Nuclear, PO Box 4000, Bowmanville, ON L1C 3Z8 (Canada)

    2008-07-15

    Ontario Power Generation Nuclear (OPGN) has a 4 x 880 MWe CANDU nuclear station at its Darlington Nuclear Div. located in Bowmanville. The station operates a Tritium Removal Facility (TRF) to reduce and maintain low tritium levels in the Moderator and Heat Transport heavy water systems of Ontario's CANDU fleet by extracting, concentrating, immobilizing and storing as a metal tritide. Minimizing tritium releases to the environment is of paramount importance to ensure that dose to the public is as low as reasonably achievable (ALARA) and to maintain credibility with the Public. Tritium is removed from the Cryogenic Distillation System to the Tritium Immobilization System (TIS) glove box via a transfer line that is protected by a rupture disc and relief valve. An overpressure event in 2003 had caused the rupture disc to blow, resulting in the release of a significant quantity of elemental tritium into the relief valve discharge line, which ties into the contaminated exhaust system. As a result of a few similar events occurring over a number of years of TRF operation, the released elemental tritium would have been converted to tritium oxide in the presence of a stagnant moist air environment in the stainless steel discharge line. A significant amount of tritium oxide hold-up in the discharge line was anticipated. To minimize any further releases to the environment, a Bubbler System was designed to remove and recover the tritium from the discharge line. This paper summarizes the results of several Bubbler recovery runs that were made over a period of a month. Approximately 3500 Ci of tritium oxide and 230 Ci of elemental tritium were removed and collected. The tritium contained in the water produced from the Bubbler system was later safely recovered in the station's downgraded D{sub 2}O clean-up and recovery system. (authors)

  10. An introduction to the National Tritium Labeling Facility

    SciTech Connect (OSTI)

    Dorsky, A.M.; Morimoto, H.; Saljoughian, M.; Williams, P.G.; Rapoport, H.

    1988-06-01

    The facilities and projects of the National Tritium Labeling Facility are described. 5 refs., 1 fig., 1 tab.

  11. The Safety and Tritium Applied Research (STAR) Facility: Status-2004

    SciTech Connect (OSTI)

    Anderl, R.A.; Longhurst, G.R.; Pawelko, R.J.; Sharpe, J.P.; Schuetz, S.T.; Petti, D.A.

    2005-07-15

    The Safety and Tritium Applied Research (STAR) Facility, a US DOE National User Facility at the Idaho National Engineering and Environmental Laboratory (INEEL), comprises capabilities and infrastructure to support both tritium and non-tritium research activities important to the development of safe and environmentally friendly fusion energy. Research thrusts include (1) interactions of tritium and deuterium with plasma-facing-component (PFC) materials, (2) fusion safety issues [PFC material chemical reactivity and dust/debris generation, activation product mobilization, tritium behavior in fusion systems], and (3) molten salts and fusion liquids for tritium breeder and coolant applications. This paper updates the status of STAR and the capabilities for ongoing research activities, with an emphasis on the development, testing and integration of the infrastructure to support tritium research activities. Key elements of this infrastructure include a tritium storage and assay system, a tritium cleanup system to process glovebox and experiment tritiated effluent gases, and facility tritium monitoring systems.

  12. Tritium Facilities Modernization and Consolidation Project Process Waste Assessment (Project S-7726)

    SciTech Connect (OSTI)

    Hsu, R.H. [Westinghouse Savannah River Company, AIKEN, SC (United States); Oji, L.N.

    1997-11-14

    Under the Tritium Facility Modernization {ampersand} Consolidation (TFM{ampersand}C) Project (S-7726) at the Savannah River Site (SS), all tritium processing operations in Building 232-H, with the exception of extraction and obsolete/abandoned systems, will be reestablished in Building 233-H. These operations include hydrogen isotopic separation, loading and unloading of tritium shipping and storage containers, tritium recovery from zeolite beds, and stripping of nitrogen flush gas to remove tritium prior to stack discharge. The scope of the TFM{ampersand}C Project also provides for a new replacement R&D tritium test manifold in 233-H, upgrading of the 233- H Purge Stripper and 233-H/234-H building HVAC, a new 234-H motor control center equipment building and relocating 232-H Materials Test Facility metallurgical laboratories (met labs), flow tester and life storage program environment chambers to 234-H.

  13. Operational Readiness Review: Savannah River Replacement Tritium Facility

    SciTech Connect (OSTI)

    Not Available

    1993-02-01

    The Operational Readiness Review (ORR) is one of several activities to be completed prior to introducing tritium into the Replacement Tritium Facility (RTF) at the Savannah River Site (SRS). The Secretary of Energy will rely in part on the results of this ORR in deciding whether the startup criteria for RTF have been met. The RTF is a new underground facility built to safely service the remaining nuclear weapons stockpile. At RTF, tritium will be unloaded from old components, purified and enriched, and loaded into new or reclaimed reservoirs. The RTF will replace an aging facility at SRS that has processed tritium for more than 35 years. RTF has completed construction and is undergoing facility startup testing. The final stages of this testing will require the introduction of limited amounts of tritium. The US Department of Energy (DOE) ORR was conducted January 19 to February 4, 1993, in accordance with an ORR review plan which was developed considering previous readiness reviews. The plan also considered the Defense Nuclear Facilities Safety Board (DNFSB) Recommendations 90-4 and 92-6, and the judgements of experienced senior experts. The review covered three major areas: (1) Plant and Equipment Readiness, (2) Personnel Readiness, and (3) Management Systems. The ORR Team was comprised of approximately 30 members consisting of a Team Leader, Senior Safety Experts, and Technical Experts. The ORR objectives and criteria were based on DOE Orders, industry standards, Institute of Nuclear Power Operations guidelines, recommendations of external oversight groups, and experience of the team members.

  14. Tritium Operation Improvements at the Idaho National Laboratory (INL) Safety and Tritium Applied Research (STAR) facility

    Broader source: Energy.gov [DOE]

    Presentation from the 35th Tritium Focus Group Meeting held in Princeton, New Jersey on May 05-07, 2015.

  15. Tritium research activities in Safety and Tritium Applied Research (STAR) facility, Idaho National Laboratory

    Office of Energy Efficiency and Renewable Energy (EERE)

    Presentation from the 32nd Tritium Focus Group Meeting held in Germantown, Maryland on April 23-25, 2013.

  16. Tritium Permeation Activity at Safety and Tritium Applied Research (STAR) Facility

    Broader source: Energy.gov [DOE]

    Presentation from the 34th Tritium Focus Group Meeting held in Idaho Falls, Idaho on September 23-25, 2014.

  17. NNSA TRITIUM SUPPLY CHAIN

    SciTech Connect (OSTI)

    Wyrick, Steven; Cordaro, Joseph; Founds, Nanette; Chambellan, Curtis

    2013-08-21

    Savannah River Site plays a critical role in the Tritium Production Supply Chain for the National Nuclear Security Administration (NNSA). The entire process includes: • Production of Tritium Producing Burnable Absorber Rods (TPBARs) at the Westinghouse WesDyne Nuclear Fuels Plant in Columbia, South Carolina • Production of unobligated Low Enriched Uranium (LEU) at the United States Enrichment Corporation (USEC) in Portsmouth, Ohio • Irradiation of TPBARs with the LEU at the Tennessee Valley Authority (TVA) Watts Bar Reactor • Extraction of tritium from the irradiated TPBARs at the Tritium Extraction Facility (TEF) at Savannah River Site • Processing the tritium at the Savannah River Site, which includes removal of nonhydrogen species and separation of the hydrogen isotopes of protium, deuterium and tritium.

  18. Report of Survey of the Los Alamos Tritium Systems Test Assembly Facility

    Broader source: Energy.gov [DOE]

    The purpose of this document is to report the results of a survey conducted at the Los Alamos Tritium Systems Test Assembly (TSTA Facility). The survey was conducted during the week of 3/20/00.

  19. Tritium research activities in Safety and Tritium Applied Research...

    Office of Environmental Management (EM)

    research activities in Safety and Tritium Applied Research (STAR) facility, Idaho National Laboratory Tritium research activities in Safety and Tritium Applied Research (STAR)...

  20. Tritium Permeation Activity at Safety and Tritium Applied Research...

    Office of Environmental Management (EM)

    Research (STAR) Facility More Documents & Publications Tritium Behavior in Lead Lithium Eutectic (LLE) at Low Tritium Partial Pressure Tritium Plasma Experiment and Its Role...

  1. Tritium monitoring in groundwater and evaluation of model predictions for the Hanford Site 200 Area Effluent Treatment Facility

    SciTech Connect (OSTI)

    Barnett, D.B.; Bergeron, M.P.; Cole, C.R.; Freshley, M.D.; Wurstner, S.K.

    1997-08-01

    The Effluent Treatment Facility (ETF) disposal site, also known as the State-Approved Land Disposal Site (SALDS), receives treated effluent containing tritium, which is allowed to infiltrate through the soil column to the water table. Tritium was first detected in groundwater monitoring wells around the facility in July 1996. The SALDS groundwater monitoring plan requires revision of a predictive groundwater model and reevaluation of the monitoring well network one year from the first detection of tritium in groundwater. This document is written primarily to satisfy these requirements and to report on analytical results for tritium in the SALDS groundwater monitoring network through April 1997. The document also recommends an approach to continued groundwater monitoring for tritium at the SALDS. Comparison of numerical groundwater models applied over the last several years indicate that earlier predictions, which show tritium from the SALDS approaching the Columbia River, were too simplified or overly robust in source assumptions. The most recent modeling indicates that concentrations of tritium above 500 pCi/L will extend, at most, no further than {approximately}1.5 km from the facility, using the most reasonable projections of ETF operation. This extent encompasses only the wells in the current SALDS tritium-tracking network.

  2. Independent Oversight Review, Savannah River Site Tritium Facilities -

    Energy Savers [EERE]

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  3. Independent Oversight Review, Savannah River Site Tritium Facilities - June

    Broader source: Energy.gov (indexed) [DOE]

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  4. Tritium monitor

    DOE Patents [OSTI]

    Chastagner, Philippe (Augusta, GA)

    1994-01-01

    A system for continuously monitoring the concentration of tritium in an aqueous stream. The system pumps a sample of the stream to magnesium-filled combustion tube which reduces the sample to extract hydrogen gas. The hydrogen gas is then sent to an isotope separation device where it is separated into two groups of isotopes: a first group of isotopes containing concentrations of deuterium and tritium, and a second group of isotopes having substantially no deuterium and tritium. The first group of isotopes containing concentrations of deuterium and tritium is then passed through a tritium detector that produces an output proportional to the concentration of tritium detected. Preferably, the detection system also includes the necessary automation and data collection equipment and instrumentation for continuously monitoring an aqueous stream.

  5. Tritium monitor

    DOE Patents [OSTI]

    Chastagner, P.

    1994-06-14

    A system is described for continuously monitoring the concentration of tritium in an aqueous stream. The system pumps a sample of the stream to magnesium-filled combustion tube which reduces the sample to extract hydrogen gas. The hydrogen gas is then sent to an isotope separation device where it is separated into two groups of isotopes: a first group of isotopes containing concentrations of deuterium and tritium, and a second group of isotopes having substantially no deuterium and tritium. The first group of isotopes containing concentrations of deuterium and tritium is then passed through a tritium detector that produces an output proportional to the concentration of tritium detected. Preferably, the detection system also includes the necessary automation and data collection equipment and instrumentation for continuously monitoring an aqueous stream. 1 fig.

  6. Tritium Behavior in Lead Lithium Eutectic (LLE) at Low Tritium...

    Office of Environmental Management (EM)

    Applied Research (STAR) Facility Tritium Plasma Experiment and Its Role in PHENIX Program Fusion Nuclear Science and Technology Program - Status and Plans for Tritium Research...

  7. An overview of research activities on materials for nuclear applications at the INL Safety, Tritium and Applied Research facility

    SciTech Connect (OSTI)

    P. Calderoni; P. Sharpe; M. Shimada

    2009-09-01

    The Safety, Tritium and Applied Research facility at the Idaho National Laboratory is a US Department of Energy National User Facility engaged in various aspects of materials research for nuclear applications related to fusion and advanced fission systems. Research activities are mainly focused on the interaction of tritium with materials, in particular plasma facing components, liquid breeders, high temperature coolants, fuel cladding, cooling and blanket structures and heat exchangers. Other activities include validation and verification experiments in support of the Fusion Safety Program, such as beryllium dust reactivity and dust transport in vacuum vessels, and support of Advanced Test Reactor irradiation experiments. This paper presents an overview of the programs engaged in the activities, which include the US-Japan TITAN collaboration, the US ITER program, the Next Generation Power Plant program and the tritium production program, and a presentation of ongoing experiments as well as a summary of recent results with emphasis on fusion relevant materials.

  8. Nuclear Material Control and Accountability (NMC&A) for the Savannah River Site Tritium Facilities

    Broader source: Energy.gov [DOE]

    Presentation from the 35th Tritium Focus Group Meeting held in Princeton, New Jersey on May 05-07, 2015.

  9. Tritium Operation Improvements at the Idaho National Laboratory...

    Office of Environmental Management (EM)

    Tritium Operation Improvements at the INL STAR facility More Documents & Publications Fusion Nuclear Science and Technology Program - Status and Plans for Tritium Research Tritium...

  10. Darlington tritium removal facility and station upgrading plant dynamic process simulation

    SciTech Connect (OSTI)

    Busigin, A. [NITEK USA, Inc., 6405 NW 77 PL, Parkland, FL 33067 (United States); Williams, G. I. D.; Wong, T. C. W.; Kulczynski, D.; Reid, A. [Ontario Power Generation Nuclear, Box 4000, Bowmanville, ON L1C 3Z8 (Canada)

    2008-07-15

    Ontario Power Generation Nuclear (OPGN) has a 4 x 880 MWe CANDU nuclear station at its Darlington Nuclear Div. located in Bowmanville. The station has been operating a Tritium Removal Facility (TRF) and a D{sub 2}O station Upgrading Plant (SUP) since 1989. Both facilities were designed with a Distributed Control System (DCS) and programmable logic controllers (PLC) for process control. This control system was replaced with a DCS only, in 1998. A dynamic plant simulator was developed for the Darlington TRF (DTRF) and the SUP, as part of the computer control system replacement. The simulator was used to test the new software, required to eliminate the PLCs. The simulator is now used for operator training and testing of process control software changes prior to field installation. Dynamic simulation will be essential for the ITER isotope separation system, where the process is more dynamic than the relatively steady-state DTRF process. This paper describes the development and application of the DTRF and SUP dynamic simulator, its benefits, architecture, and the operational experience with the simulator. (authors)

  11. Sources of tritium

    SciTech Connect (OSTI)

    Phillips, J.E.; Easterly, C.E.

    1980-12-01

    A review of tritium sources is presented. The tritium production and release rates are discussed for light water reactors (LWRs), heavy water reactors (HWRs), high temperature gas cooled reactors (HTGRs), liquid metal fast breeder reactors (LMFBRs), and molten salt breeder reactors (MSBRs). In addition, release rates are discussed for tritium production facilities, fuel reprocessing plants, weapons detonations, and fusion reactors. A discussion of the chemical form of the release is included. The energy producing facilities are ranked in order of increasing tritium production and release. The ranking is: HTGRs, LWRs, LMFBRs, MSBRs, and HWRs. The majority of tritium has been released in the form of tritiated water.

  12. ANALYSIS OF THE TRITIUM-WATER (T-H20) SYSTEM FOR A FUSION MATERIAL TEST FACILITY

    E-Print Network [OSTI]

    Harilal, S. S.

    ....................................... 2 3. Thermal Hydraulics and Vacuum System Requirements ............. 6 4. Cost of Tritium Recovery

  13. Facility effluent monitoring plan for the Plutonium Uranium Extraction Facility

    SciTech Connect (OSTI)

    Greager, E.M.

    1997-12-11

    A facility effluent monitoring plan is required by the US Department of Energy in DOE Order 5400.1 for any operations that involve hazardous materials and radioactive substances that could impact employee or public safety or the environment. This document is prepared using the specific guidelines identified in A Guide for Preparing Hanford Site Facility Effluent Monitoring Plans, WHC-EP-0438-01. This facility effluent monitoring plan assesses effluent monitoring systems and evaluates whether these systems are adequate to ensure the public health and safety as specified in applicable federal, state, and local requirements. This facility effluent monitoring plan will ensure long-range integrity of the effluent monitoring systems by requiring an update whenever a new process or operation introduces new hazardous materials or significant radioactive materials. This document must be reviewed annually even if there are no operational changes, and it must be updated, at a minimum, every 3 years.

  14. Tritium monitoring techniques

    SciTech Connect (OSTI)

    DeVore, J.R.; Buckner, M.A.

    1996-05-01

    As part of their operations, the U.S. Navy is required to store or maintain operational nuclear weapons on ships and at shore facilities. Since these weapons contain tritium, there are safety implications relevant to the exposure of personnel to tritium. This is particularly important for shipboard operations since these types of environments can make low-level tritium detection difficult. Some of these ships have closed systems, which can result in exposure to tritium at levels that are below normally acceptable levels but could still cause radiation doses that are higher than necessary or could hamper ship operations. This report describes the state of the art in commercial tritium detection and monitoring and recommends approaches for low-level tritium monitoring in these environments.

  15. Corrosion within the Z-Bed Recovery Systems at the Savannah River Site’s Tritium Facilities

    Broader source: Energy.gov [DOE]

    Presentation from the 32nd Tritium Focus Group Meeting held in Germantown, Maryland on April 23-25, 2013.

  16. Mobility of Tritium in Engineered and Earth Materials at the NuMI Facility, Fermilab: Progress report for work performed between June 13 and September 30, 2006

    E-Print Network [OSTI]

    2006-01-01

    Drainage System The water and tritium observed at the sump (sump holding tank is affected by the tritium transport mechanisms from the tritium source zones to the drainage

  17. Demonstration of High Performance in Layered Deuterium-Tritium Capsule Implosions in Uranium Hohlraums at the National Ignition Facility

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Döppner, T.; Callahan, D. A.; Hurricane, O. A.; Hinkel, D. E.; Ma, T.; Park, H. -S.; Berzak Hopkins, L. F.; Casey, D. T.; Celliers, P. P.; Dewald, E. L.; et al

    2015-07-28

    We report on the first layered deuterium-tritium (DT) capsule implosions indirectly driven by a “highfoot” laser pulse that were fielded in depleted uranium hohlraums at the National Ignition Facility. Recently, high-foot implosions have demonstrated improved resistance to ablation-front Rayleigh-Taylor instability induced mixing of ablator material into the DT hot spot [Hurricane et al., Nature (London) 506, 343 (2014)]. Uranium hohlraums provide a higher albedo and thus an increased drive equivalent to an additional 25 TW laser power at the peak of the drive compared to standard gold hohlraums leading to higher implosion velocity. Additionally, we observe an improved hot-spot shapemore »closer to round which indicates enhanced drive from the waist. In contrast to findings in the National Ignition Campaign, now all of our highest performing experiments have been done in uranium hohlraums and achieved total yields approaching 1016 neutrons where more than 50% of the yield was due to additional heating of alpha particles stopping in the DT fuel.« less

  18. Storage and Assay of Tritium in STAR

    SciTech Connect (OSTI)

    Longhurst, Glen R.; Anderl, Robert A.; Pawelko, Robert J.; Stoots, Carl J.

    2005-07-15

    The Safety and Tritium Applied Research (STAR) facility at the Idaho National Engineering and Environmental Laboratory (INEEL) is currently being commissioned to investigate tritium-related safety questions for fusion and other technologies. The tritium inventory for the STAR facility will be maintained below 1.5 g to avoid the need for STAR to be classified as a Category 3 nuclear facility. A key capability in successful operation of the STAR facility is the ability to receive, inventory, and dispense tritium to the various experiments underway there. The system central to that function is the Tritium Storage and Assay System (SAS).The SAS has four major functions: (1) receiving and holding tritium, (2) assaying, (3) dispensing, and (4) purifying hydrogen isotopes from non-hydrogen species.This paper describes the design and operation of the STAR SAS and the procedures used for tritium accountancy in the STAR facility.

  19. Construction and Operation of a Tritium Extraction Facility at the Savannah Siver Site

    National Nuclear Security Administration (NNSA)

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  20. Tritium breeding materials

    SciTech Connect (OSTI)

    Hollenberg, G.W.; Johnson, C.E.; Abdou, M.

    1984-03-01

    Tritium breeding materials are essential to the operation of D-T fusion facilities. Both of the present options - solid ceramic breeding materials and liquid metal materials are reviewed with emphasis not only on their attractive features but also on critical materials issues which must be resolved.

  1. Storage and Assay of Tritium in STAR

    SciTech Connect (OSTI)

    Glen R. Longhurst; Robert A. Anderl; Robert J. Pawelko

    2004-09-01

    The Safety and Tritium Applied Research (STAR) facility has recently been commissioned to investigate tritium-related safety questions for fusion and other technologies. The authorized inventory of tritium is 1.6 grams, the threshold quantity for nuclear facility classification. A key capability in successful operation of the STAR facility is the ability to receive, inventory, and dispense tritium to the various experiments underway there. The system central to that function is the Tritium Storage and Assay System (SAS). The SAS has four major functions: (1) receiving and holding tritium from shipping containers brought into the STAR facility, (2) assaying the amount of tritium in the SAS, (3) dispensing tritium to secondary beds or containers used for transferring it to the experimental systems in the STAR facility, and (4) purifying hydrogen isotopes from non-hydrogen species. To that may be added a fifth, optional function, isotopic separation of hydrogen isotopes using bed-to-bed transfer techniques. This paper documents the design and operation of the STAR SAS and the procedures used for tritium accountancy in the STAR facility.

  2. Tritium systems test assembly stabilization

    SciTech Connect (OSTI)

    Jasen, W. G. (William G.); Michelotti, R. A. (Roy A.); Anast, K. R. (Kurt R.); Tesch, Charles

    2004-01-01

    The Tritium Systems Test Assembly (TSTA) was a facility dedicated to tritium technology Research and Development (R&D) primarily for future fusion power reactors. The facility was conceived in mid 1970's, operations commenced in early 1980's, stabilization and deactivation began in 2000 and were completed in 2003. The facility will remain in a Surveillance and Maintenance (S&M) mode until the Department of Energy (DOE) funds demolition of the facility, tentatively in 2009. A safe and stable end state was achieved by the TSTA Facility Stabilization Project (TFSP) in anticipation of long term S&M. At the start of the stabilization project, with an inventory of approximately 140 grams of tritium, the facility was designated a Hazard Category (HC) 2 Non-Reactor Nuclear facility as defined by US Department of Energy standard DOE-STD-1027-92 (1997). The TSTA facility comprises a laboratory area, supporting rooms, offices and associated laboratory space that included more than 20 major tritium handling systems. The project's focus was to reduce the tritium inventory by removing bulk tritium, tritiated water wastes, and tritium-contaminated high-inventory components. Any equipment that remained in the facility was stabilized in place. All of the gloveboxes and piping were rendered inoperative and vented to atmosphere. All equipment, and inventoried tritium contamination, remaining in the facility was left in a safe-and-stable state. The project used the End Points process as defined by the DOE Office of Environmental Management (web page http://www.em.doe.- gov/deact/epman.htmtlo) document and define the end state required for the stabilization of TSTA Facility. The End Points process added structure that was beneficial through virtually all phases of the project. At completion of the facility stabilization project the residual tritium inventory was approximately 3,000 curies, considerably less than the 1.6-gram threshold for a HC 3 facility. TSTA is now designated as a Radiological Facility. Innovative approaches were employed for characterization and removal of legacy wastes and high inventory components. Major accomplishments included: (1) Reduction of tritium inventory, elimination of chemical hazards, and identification and posting of remaining hazards. (2) Removal of legacy wastes. (3) Transferred equipment for reuse in other DOE projects, including some at other DOE facilities. (4) Transferred facility in a safe and stable condition to the S&M organization. The project successfully completed all project goals and the TSTA facility was transferred into S&M on August 1,2003. This project demonstrates the benefit of radiological inventory reduction and the removal of legacy wastes to achieve a safe and stable end state that protects workers and the environment pending eventual demolition of the facility.

  3. Use of system code to estimate equilibrium tritium inventory in fusion DT machines, such as ARIES-AT and components testing facilities

    SciTech Connect (OSTI)

    C.P.C. Wong; B. Merrill

    2014-10-01

    ITER is under construction and will begin operation in 2020. This is the first 500 MWfusion class DT device, and since it is not going to breed tritium, it will consume most of the limited supply of tritium resources in the world. Yet, in parallel, DT fusion nuclear component testing machines will be needed to provide technical data for the design of DEMO. It becomes necessary to estimate the tritium burn-up fraction and corresponding initial tritium inventory and the doubling time of these machines for the planning of future supply and utilization of tritium. With the use of a system code, tritium burn-up fraction and initial tritium inventory for steady state DT machines can be estimated. Estimated tritium burn-up fractions of FNSF-AT, CFETR-R and ARIES-AT are in the range of 1–2.8%. Corresponding total equilibrium tritium inventories of the plasma flow and tritium processing system, and with the DCLL blanket option are 7.6 kg, 6.1 kg, and 5.2 kg for ARIES-AT, CFETR-R and FNSF-AT, respectively.

  4. Tritium handling in vacuum systems

    SciTech Connect (OSTI)

    Gill, J.T. [Monsanto Research Corp., Miamisburg, OH (United States). Mound Facility; Coffin, D.O. [Los Alamos National Lab., NM (United States)

    1986-10-01

    This report provides a course in Tritium handling in vacuum systems. Topics presented are: Properties of Tritium; Tritium compatibility of materials; Tritium-compatible vacuum equipment; and Tritium waste treatment.

  5. EA-0874: Low-level Waste Drum Staging Building at Weapons Engineering Tritium Facility, TA-16 Los Alamos National Laboratory, Los Alamos, New Mexico

    Broader source: Energy.gov [DOE]

    This EA evaluates the environmental impacts of a proposal to place a 3 meter (m) by 4.5 m prefabricated storage building (transportainer) adjacent to the existing Weapons Engineering Tritium...

  6. Pf/Zeolite Catalyst for Tritium Stripping

    SciTech Connect (OSTI)

    Hsu, R.H.

    2001-03-26

    This report described promising hydrogen (protium and tritium) stripping results obtained with a Pd/zeolite catalyst at ambient temperature. Preliminary results show 90-99+ percent tritium stripping efficiency may be obtained, with even better performance expected as bed configuration and operating conditions are optimized. These results suggest that portable units with single beds of the Pd/zeolite catalyst may be utilized as ''catalytic absorbers'' to clean up both tritium gas and tritiated water. A cart-mounted prototype stripper utilizing this catalyst has been constructed for testing. This portable stripper has potential applications in maintenance-type jobs such as tritium line breaks. This catalyst can also potentially be utilized in an emergency stripper for the Replacement Tritium Facility.

  7. Disposal of tritium residues at the Los Alamos National Laboratory. Audit repost

    SciTech Connect (OSTI)

    NONE

    1998-07-01

    The objective of this audit was to determine whether Los Alamos disposed of wastewater containing tritium residues in a safe and cost-effective manner subsequent to an October 1991 report reviewing tritium facility management practices.

  8. Tritium Focus Group Meeting:

    Office of Environmental Management (EM)

    void etc.) in bulk PFCs - Large amount of tritium can be trapped in vacancy-cluster as gas form, leading to bubble formation, and blister formation in metal - Tritium behavior in...

  9. Radiological Training for Tritium Facilities

    Energy Savers [EERE]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on DeliciousMathematicsEnergyInterested PartiesBuildingBudget ||Department ofRequest7of 9 Radiological Control8

  10. Radiological Training for Tritium Facilities

    Energy Savers [EERE]

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on DeliciousMathematicsEnergyInterested PartiesBuildingBudget ||Department ofRequest7of 9 Radiological Control8DOE

  11. Continuous aqueous tritium monitor

    DOE Patents [OSTI]

    McManus, G.J.; Weesner, F.J.

    1987-10-19

    An apparatus for a selective on-line determination of aqueous tritium concentration is disclosed. A moist air stream of the liquid solution being analyzed is passed through a permeation dryer where the tritium and moisture are selectively removed to a purge air stream. The purge air stream is then analyzed for tritium concentration, humidity, and temperature, which allows computation of liquid tritium concentration. 2 figs.

  12. Tritium 2013 Presentation

    Broader source: Energy.gov [DOE]

    Presentation from the 32nd Tritium Focus Group Meeting held in Germantown, Maryland on April 23-25, 2013.

  13. Tritium Focus Group- INEL

    Broader source: Energy.gov [DOE]

    Presentation from the 34th Tritium Focus Group Meeting held in Idaho Falls, Idaho on September 23-25, 2014.

  14. Introduction Airborne Tritium Tritides

    Broader source: Energy.gov [DOE]

    Presentation from the 33rd Tritium Focus Group Meeting held in Aiken, South Carolina on April 22-24, 2014.

  15. Tritium on Metal Surfaces

    Broader source: Energy.gov [DOE]

    Presentation from the 34th Tritium Focus Group Meeting held in Idaho Falls, Idaho on September 23-25, 2014.

  16. Bulk Tritium Shipping Package

    Broader source: Energy.gov [DOE]

    Presentation from the 32nd Tritium Focus Group Meeting held in Germantown, Maryland on April 23-25, 2013.

  17. PDRD (SR13046) TRITIUM PRODUCTION FINAL REPORT

    SciTech Connect (OSTI)

    Smith, P.; Sheetz, S.

    2013-09-30

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

  18. For economic energy, we need: tritium, large size to obtain hot fusing plasma; high fields and large currents

    E-Print Network [OSTI]

    11 For economic energy, we need: tritium, large size to obtain hot fusing plasma; high fields: a Component Test Facility is much needed; ST appears simplest and most economic in tritium: BUT the high cost

  19. Tritium Technology at CNL

    Office of Environmental Management (EM)

    ILLIMIT * Use of tritium decay-energy as power source * Convert high-energy electrons into low-voltage current * Two methods: * Indirect: beta photons conduction...

  20. Oxidative Tritium Decontamination System

    DOE Patents [OSTI]

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

    2006-02-07

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

  1. Microsoft Word - Tritium Production and Environmental Impacts...

    National Nuclear Security Administration (NNSA)

    be released to the environment. The operational differences between a tritium production reactor and a nuclear power plant without tritium production are as follows: * Tritium...

  2. Tritium catalyzed deuterium tokamaks

    SciTech Connect (OSTI)

    Greenspan, E.; Miley, G.H.; Jung, J.; Gilligan, J.

    1984-04-01

    A preliminary assessment of the promise of the Tritium Catalyzed Deuterium (TCD) tokamak power reactors relative to that of deuterium-tritium (D-T) and catalyzed deuterium (Cat-D) tokamaks is undertaken. The TCD mode of operation is arrived at by converting the /sup 3/He from the D(D,n)/sup 3/He reaction into tritium, by neutron capture in the blanket; the tritium thus produced is fed into the plasma. There are three main parts to the assessment: blanket study, reactor design and economic analysis and an assessment of the prospects for improvements in the performance of TCD reactors (and in the promise of the TCD mode of operation, in general).

  3. Tritium waste package

    DOE Patents [OSTI]

    Rossmassler, R.; Ciebiera, L.; Tulipano, F.J.; Vinson, S.; Walters, R.T.

    1995-11-07

    A containment and waste package system for processing and shipping tritium oxide waste received from a process gas includes an outer drum and an inner drum containing a disposable molecular sieve bed (DMSB) seated within the outer drum. The DMSB includes an inlet diffuser assembly, an outlet diffuser assembly, and a hydrogen catalytic recombiner. The DMSB absorbs tritium oxide from the process gas and converts it to a solid form so that the tritium is contained during shipment to a disposal site. The DMSB is filled with type 4A molecular sieve pellets capable of adsorbing up to 1000 curies of tritium. The recombiner contains a sufficient amount of catalyst to cause any hydrogen and oxygen present in the process gas to recombine to form water vapor, which is then adsorbed onto the DMSB. 1 fig.

  4. Wet processing of palladium for use in the tritium facility at Westinghouse, Savannah River, SC. Preparation of palladium using the Mound Muddy Water process

    SciTech Connect (OSTI)

    Baldwin, D.P.; Zamzow, D.S.

    1998-11-10

    Palladium used at Savannah River for tritium storage is currently obtained from a commercial source. In order to better understand the processes involved in preparing this material, Savannah River is supporting investigations into the chemical reactions used to synthesize this material and into the conditions necessary to produce palladium powder that meets their specifications. This better understanding may help to guarantee a continued reliable source for this material in the future. As part of this evaluation, a work-for-others contract between Westinghouse Savannah River Company and the Ames Laboratory Metallurgy and Ceramics Program was initiated. During FY98, the process for producing palladium powder developed in 1986 by Dan Grove of Mound Applied Technologies (USDOE) was studied to understand the processing conditions that lead to changes in morphology in the final product. This report details the results of this study of the Mound Muddy Water process, along with the results of a round-robin analysis of well-characterized palladium samples that was performed by Savannah River and Ames Laboratory. The Mound Muddy Water process is comprised of three basic wet chemical processes, palladium dissolution, neutralization, and precipitation, with a number of filtration steps to remove unwanted impurity precipitates.

  5. Tritium Permeation Activity at Safety and Tritium Applied

    Office of Environmental Management (EM)

    safety exp. in Tritium Area * Size: 4300 ft 2 400 m 2 3 Safety concern: ex-vessel release source term * Challenges in blanket development - Tritium permeation leads to the...

  6. Measurements of collective fuel velocities in deuterium-tritium exploding pusher and cryogenically layered deuterium-tritium implosions on the NIF

    E-Print Network [OSTI]

    Measurements of collective fuel velocities in deuterium-tritium exploding pusher and cryogenically://pop.aip.org/about/rights_and_permissions #12;Measurements of collective fuel velocities in deuterium-tritium exploding pusher and cryogenically fuel velocities in Inertial Confinement Fusion implosions at the National Ignition Facility

  7. Experience with Palladium Diffusers in Tritium Processing

    SciTech Connect (OSTI)

    Motyka, T.; Clark, E.A.; Dauchess, D.A.; Heung, L.K.; Rabum, R.L.

    1995-01-27

    Hydrogen isotopes are separated from other gases by permeation through palladium and palladium-silver alloy diffusers in the Tritium Facilities at the US Department of Energy Savannah River Site (SRS). Diffusers have provided effective service for almost forty years. This paper is an overview of the operational experience with the various diffuser types that have been employed at SRS. Alternative technologies being developed at SRS for purifying hydrogen isotopes are also discussed.

  8. Tritium Effects on Reservoir Materials

    Broader source: Energy.gov [DOE]

    Presentation from the 32nd Tritium Focus Group Meeting held in Germantown, Maryland on April 23-25, 2013.

  9. Tritium Design Practices: Part 2

    Broader source: Energy.gov [DOE]

    Presentation from the 32nd Tritium Focus Group Meeting held in Germantown, Maryland on April 23-25, 2013.

  10. Advanced Polymers for Tritium Service

    Broader source: Energy.gov [DOE]

    Presentation from the 34th Tritium Focus Group Meeting held in Idaho Falls, Idaho on September 23-25, 2014.

  11. Tritium Instrument Demonstration Station (TIDS)

    Broader source: Energy.gov [DOE]

    Presentation from the 34th Tritium Focus Group Meeting held in Idaho Falls, Idaho on September 23-25, 2014.

  12. Tritium Instrument Demonstration Station (TIDS)

    Office of Energy Efficiency and Renewable Energy (EERE)

    Presentation from the 35th Tritium Focus Group Meeting held in Princeton, New Jersey on May 05-07, 2015.

  13. Tritium Detection Methods and Limitations

    Broader source: Energy.gov [DOE]

    Presentation from the 33rd Tritium Focus Group Meeting held in Aiken, South Carolina on April 22-24, 2014.

  14. Tritium Instrument Demonstration Station (TIDS)

    Broader source: Energy.gov [DOE]

    Presentation from the 33rd Tritium Focus Group Meeting held in Aiken, South Carolina on April 22-24, 2014.

  15. MODELING ATMOSPHERIC RELEASES OF TRITIUM FROM NUCLEAR INSTALLATIONS

    SciTech Connect (OSTI)

    Okula, K

    2007-01-17

    Tritium source term analysis and the subsequent dispersion and consequence analyses supporting the safety documentation of Department of Energy nuclear facilities are especially sensitive to the applied software analysis methodology, input data and user assumptions. Three sequential areas in tritium accident analysis are examined in this study to illustrate where the analyst should exercise caution. Included are: (1) the development of a tritium oxide source term; (2) use of a full tritium dispersion model based on site-specific information to determine an appropriate deposition scaling factor for use in more simplified, broader modeling, and (3) derivation of a special tritium compound (STC) dose conversion factor for consequence analysis, consistent with the nature of the originating source material. It is recommended that unless supporting, defensible evidence is available to the contrary, the tritium release analyses should assume tritium oxide as the species released (or chemically transformed under accident's environment). Important exceptions include STC situations and laboratory-scale releases of hydrogen gas. In the modeling of the environmental transport, a full phenomenology model suggests that a deposition velocity of 0.5 cm/s is an appropriate value for environmental features of the Savannah River Site. This value is bounding for certain situations but non-conservative compared to the full model in others. Care should be exercised in choosing other factors such as the exposure time and the resuspension factor.

  16. 2009 EVALUATION OF TRITIUM REMOVAL AND MITIGATION TECHNOLOGIES FOR WASTEWATER TREATMENT

    SciTech Connect (OSTI)

    LUECK KJ; GENESSE DJ; STEGEN GE

    2009-02-26

    Since 1995, a state-approved land disposal site (SALDS) has received tritium contaminated effluents from the Hanford Site Effluent Treatment Facility (ETF). Tritium in this effluent is mitigated by storage in slow moving groundwater to allow extended time for decay before the water reaches the site boundary. By this method, tritium in the SALDS is isolated from the general environment and human contact until it has decayed to acceptable levels. This report contains the 2009 update evaluation of alternative tritium mitigation techniques to control tritium in liquid effluents and groundwater at the Hanford site. A thorough literature review was completed and updated information is provided on state-of-the-art technologies for control of tritium in wastewaters. This report was prepared to satisfy the Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) Milestone M-026-07B (Ecology, EPA, and DOE 2007). Tritium separation and isolation technologies are evaluated periodically to determine their feasibility for implementation to control Hanford site liquid effluents and groundwaters to meet the Us. Code of Federal Regulations (CFR), Title 40 CFR 141.16, drinking water maximum contaminant level (MCL) for tritium of 20,000 pOll and/or DOE Order 5400.5 as low as reasonably achievable (ALARA) policy. Since the 2004 evaluation, there have been a number of developments related to tritium separation and control with potential application in mitigating tritium contaminated wastewater. These are primarily focused in the areas of: (1) tritium recycling at a commercial facility in Cardiff, UK using integrated tritium separation technologies (water distillation, palladium membrane reactor, liquid phase catalytic exchange, thermal diffusion), (2) development and demonstration of Combined Electrolysis Catalytic Exchange (CECE) using hydrogen/water exchange to separate tritium from water, (3) evaporation of tritium contaminated water for dispersion in the atmosphere, and (4) use of barriers to minimize the transport of tritium in groundwater. Continuing development efforts for tritium separations processes are primarily to support the International Thermonuclear Experimental Reactor (ITER) program, the nuclear power industry, and the production of radiochemicals. While these applications are significantly different than the Hanford application, the technology could potentially be adapted for Hanford wastewater treatment. Separations based processes to reduce tritium levels below the drinking water MCL have not been demonstrated for the scale and conditions required for treating Hanford wastewater. In addition, available cost information indicates treatment costs for such processes will be substantially higher than for discharge to SALDS or other typical pump and treat projects at Hanford. Actual mitigation projects for groundwater with very low tritium contamination similar to that found at Hanford have focused mainly on controlling migration and on evaporation for dispersion in the atmosphere.

  17. Drum bubbler tritium processing system

    DOE Patents [OSTI]

    Rule, K.; Gettelfinger, G.; Kivler, P.

    1999-08-17

    A method is described for separating tritium oxide from a gas stream containing tritium oxide. The gas stream containing tritium oxide is fed into a container of water having a head space above the water. The tritium oxide is separated by bubbling the gas stream containing tritium oxide through the container of water and removing gas from the container head space above the water. Thereafter, the gas from the head space is dried to remove water vapor from the gas, and the water vapor is recycled to the container of water. 2 figs.

  18. Tritium retention in TFTR

    SciTech Connect (OSTI)

    Dylla, H.F.; Wilson, K.L. (eds.)

    1988-04-01

    This report discusses the materials physics related to D-T operation in TFTR. Research activities are described pertaining to basic studies of hydrogenic retention in graphite, hydrogen recycling phenomena, first-wall and limiter conditioning, surface analysis of TFTR first-wall components, and estimates of the tritium inventory.

  19. PLUTONIUM-URANIUM EXTRACTION (PUREX) FACILITY ALARACT DEMONSTRATION FOR FILTER HOUSING

    SciTech Connect (OSTI)

    LEBARON GJ

    2008-11-25

    This document presents an As Low As Reasonably Achievable Control Technology (ALARACT) demonstration for evaluating corrosion on the I-beam supporting filter housing No.9 for the 291-A-l emission unit of the Plutonium-Uranium Extraction (PUREX) Facility, located in the 200 East Area of the Hanford Site. The PUREX facility is currently in surveillance and maintenance mode. During a State of Washington, Department of Health (WDOH) 291-A-l emission unit inspection, a small amount of corrosion was observed at the base of a high-efficiency particulate air (HEPA) filter housing. A series of internal and external inspections identified the source of the corrosion material as oxidation of a small section of one of the carbon steel I-beams that provides support to the stainless steel filter housing. The inspections confirmed the corrosion is isolated to one I-beam support location and does not represent any compromise of the structural support or filter housing integrity. Further testing and inspections of the support beam corrosion and its cause were conducted but did not determine the cause. No definitive evidence was found to support any degradation of the housing. Although no degradation of the housing was found, a conservative approach will be implemented. The following actions will be taken: (1) The current operating filter housing No.9 will be removed from service. (2) The only remaining available filter housings (No.1, No.2, and No.3) will be placed in service. These filter housings have new HEPA filters fitted with stainless steel frames and faceguards which were installed in the spring of 2007. (3) Filter housings No.5 and No.10 will be put on standby as backups. To document the assessment of the unit, a draft ALARACT filter housing demonstration for the PUREX filter housing was prepared, and informally provided to WDOH on August 7, 2008. A follow up WDOH response to the draft ALARACT filter housing demonstration for the PUREX filter housing questioned whether deteriorated galvanized filter faceguards discovered during an internal filter housing inspection met American Society of Mechanical Engineers (ASME) AG-l or Military Specification (MIL) 51068 standards. The filter system was designed and installed prior to the issuance of AG-l, February 1986; however, MIL 51068 did require galvanized faceguards. The faceguards are not necessary for filtration or structural purposes; it is concluded that the system is in compliance with the intent of the applicable standard. Appendix B provides supporting information to address this issue.

  20. COIIF-840 4137--1 TRITIUM BREEDING MATERIALS D E 8 4 010521

    E-Print Network [OSTI]

    Abdou, Mohamed

    breeding materials, since our limited experience with these new (and sometimes exotic) materials providesNOTICE 3b ''ity# COIIF-840 4137--1 TRITIUM BREEDING MATERIALS D E 8 4 010521 G. W. Hollenberg at LosAngeles ABSTRACT Tritium breeding materials areessential to theoperation of D-Tfusion facilities

  1. Tritium monitor and collection system

    DOE Patents [OSTI]

    Bourne, Gary L. (Idaho Falls, ID); Meikrantz, David H. (Idaho Falls, ID); Ely, Walter E. (Los Alamos, NM); Tuggle, Dale G. (Los Alamos, NM); Grafwallner, Ervin G. (Arco, ID); Wickham, Keith L. (Idaho Falls, ID); Maltrud, Herman R. (Los Alamos, NM); Baker, John D. (Blackfoot, ID)

    1992-01-01

    This system measures tritium on-line and collects tritium from a flowing inert gas stream. It separates the tritium from other non-hydrogen isotope contaminating gases, whether radioactive or not. The collecting portion of the system is constructed of various zirconium alloys called getters. These alloys adsorb tritium in any of its forms at one temperature and at a higher temperature release it as a gas. The system consists of four on-line getters and heaters, two ion chamber detectors, two collection getters, and two guard getters. When the incoming gas stream is valved through the on-line getters, 99.9% of it is adsorbed and the remainder continues to the guard getter where traces of tritium not collected earlier are adsorbed. The inert gas stream then exits the system to the decay chamber. Once the on-line getter has collected tritium for a predetermined time, it is valved off and the next on-line getter is valved on. Simultaneously, the first getter is heated and a pure helium purge is employed to carry the tritium from the getter. The tritium loaded gas stream is then routed through an ion chamber which measures the tritium activity. The ion chamber effluent passes through a collection getter that readsorbs the tritium and is removable from the system once it is loaded and is then replaced with a clean getter. Prior to removal of the collection getter, the system switches to a parallel collection getter. The effluent from the collection getter passes through a guard getter to remove traces of tritium prior to exiting the system. The tritium loaded collection getter, once removed, is analyzed by liquid scintillation techniques. The entire sequence is under computer control except for the removal and analysis of the collection getter.

  2. Tritium monitor and collection system

    DOE Patents [OSTI]

    Bourne, G.L.; Meikrantz, D.H.; Ely, W.E.; Tuggle, D.G.; Grafwallner, E.G.; Wickham, K.L.; Maltrud, H.R.; Baker, J.D.

    1992-01-14

    This system measures tritium on-line and collects tritium from a flowing inert gas stream. It separates the tritium from other non-hydrogen isotope contaminating gases, whether radioactive or not. The collecting portion of the system is constructed of various zirconium alloys called getters. These alloys adsorb tritium in any of its forms at one temperature and at a higher temperature release it as a gas. The system consists of four on-line getters and heaters, two ion chamber detectors, two collection getters, and two guard getters. When the incoming gas stream is valved through the on-line getters, 99.9% of it is adsorbed and the remainder continues to the guard getter where traces of tritium not collected earlier are adsorbed. The inert gas stream then exits the system to the decay chamber. Once the on-line getter has collected tritium for a predetermined time, it is valved off and the next on-line getter is valved on. Simultaneously, the first getter is heated and a pure helium purge is employed to carry the tritium from the getter. The tritium loaded gas stream is then routed through an ion chamber which measures the tritium activity. The ion chamber effluent passes through a collection getter that readsorbs the tritium and is removable from the system once it is loaded and is then replaced with a clean getter. Prior to removal of the collection getter, the system switches to a parallel collection getter. The effluent from the collection getter passes through a guard getter to remove traces of tritium prior to exiting the system. The tritium loaded collection getter, once removed, is analyzed by liquid scintillation techniques. The entire sequence is under computer control except for the removal and analysis of the collection getter. 7 figs.

  3. Tritium Ground Water Issues | Department of Energy

    Office of Environmental Management (EM)

    Ground Water Issues Tritium Ground Water Issues Presentation from the 35th Tritium Focus Group Meeting held in Princeton, New Jersey on May 05-07, 2015. Tritium Ground Water Issues...

  4. Particulate Generation in Tritium Systems

    Office of Environmental Management (EM)

    after 10 years of operation. * Tritium processing interrupted when other filters (flow orifices) became plugged approximately two weeks later * A fine particulate was found...

  5. Introduction to Airborne Tritium Tritides

    Office of Environmental Management (EM)

    in four ways: 1. By cosmic rays striking the earth's atmosphere. 2. By nuclear fission - the primary source. 3. Accelerator produced Tritium 4. Neutron activation and...

  6. Tritium and plutonium production as a step toward ICF commercialization

    SciTech Connect (OSTI)

    Pendergrass, J.H.; Dudziak, D.J.

    1983-01-01

    The feasibility of a combined special nuclear materials (SNM) production plant/engineering test facility (ETF) with reduced pellet and driver performance requirements as a step toward commercialization of inertial confinement fusion (ICF) is examined. Blanket design and tritium production cost studies, the status of R and D programs, and the ETF role are emphasized.

  7. Brookhaven National Laboratory - HFBR Tritium | Department of...

    Office of Environmental Management (EM)

    HFBR Tritium Brookhaven National Laboratory - HFBR Tritium January 1, 2014 - 12:00pm Addthis US Department of Energy Groundwater Database Groundwater Master Report...

  8. Design of a crystal extraction facility in the east utility straight

    SciTech Connect (OSTI)

    Dukes, E.C.; Murphy, C.T.; Parker, B.

    1993-09-01

    Parasitic extraction of a small fraction of the 20-TeV circulating beam of the Superconducting Super Collider can be done using a bent crystal situated in the halo of the orbiting beam. The authors present a design of a crystal extraction system that is compatible with current plans for momentum scraping in the east utility straight. The only modification to the collider tunnel is the addition of a 160-m-long alcove in the east utility straight to mate the extracted beam line microtunnel with the collider tunnel. No other changes to the east utility straight tunnel are needed.

  9. Tritium monitoring of groundwater and surfaces

    SciTech Connect (OSTI)

    MacArthur, D.; Aamodt, P.; Bounds, J.; Koster, J.

    1999-03-01

    There are numerous facilities, both within the US and in the rest of the world, within the complex of radiation laboratories and production plants where tritium has been released into the environment because of historic or ongoing mission-related operations. Many of environmental restoration projects have detected low levels of tritium contamination in local streams, ponds, and/or ground water. Typically these waters are moving or have the potential to move offsite and are viewed as a potential risk to the public and environment. Los Alamos National Laboratory will modify the well-proven long-range alpha detection (LRAD) technique for detection of ionizing radiation to optimize a system for detecting tritium in groundwater and other surfaces. The LRAD technique relies on detection of ionized air molecules rather than direct detection of ionizing radiation. The detected electrical current is proportional to the number of ionized air molecules present, which is in turn a measure of the amount of contamination present. Although this technique has been used commercially to measure alpha contamination on objects and surfaces, the technique is also ideal for monitoring low-energy beta particles. The authors have demonstrated beta detection using {sup 54}Mn, {sup 14}C, {sup 147}Pm, {sup 99}Tc, {sup 90}Sr, and {sup 36}Cl sources. Thus, the detector technology and detection of beta particles using this technology have both been demonstrated. The extreme short range of tritium beta particles necessitates an optimization of the detector system. In this paper, the authors will discuss these new designs.

  10. Tritium Supply Considerations Scott Willms

    E-Print Network [OSTI]

    history of US tritium production · 1953-1955 Tritium producing reactors online · 1976-1988 Need for new selected · 1991 Arms reduction progress, only one option needed K Reactor leaks #12;Brief history of US to DOE with Watts Bar/Sequoya service as backup 1998 "Interagency review" issued Watts Bar service chosen

  11. TRITIUM ACCOUNTANCY IN FUSION SYSTEMS

    SciTech Connect (OSTI)

    Klein, J. E.; Farmer, D. A.; Moore, M. L.; Tovo, L. L.; Poore, A. S.; Clark, E. A.; Harvel, C. D.

    2014-03-06

    The US Department of Energy (DOE) has clearly defined requirements for nuclear material control and accountability (MC&A) of tritium whereas the International Atomic Energy Agency (IAEA) does not since tritium is not a fissile material. MC&A requirements are expected for tritium fusion machines and will be dictated by the host country or regulatory body where the machine is operated. Material Balance Areas (MBAs) are defined to aid in the tracking and reporting of nuclear material movements and inventories. Material subaccounts (MSAs) are established along with key measurement points (KMPs) to further subdivide a MBA to localize and minimize uncertainties in the inventory difference (ID) calculations for tritium accountancy. Fusion systems try to minimize tritium inventory which may require continuous movement of material through the MSAs. The ability of making meaningful measurements of these material transfers is described in terms of establishing the MSA structure to perform and reconcile ID calculations. For fusion machines, changes to the traditional ID equation will be discussed which includes breading, burn-up, and retention of tritium in the fusion device. The concept of “net” tritium quantities consumed or lost in fusion devices is described in terms of inventory taking strategies and how it is used to track the accumulation of tritium in components or fusion machines.

  12. Fermilab | Tritium at Fermilab

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity ofkandz-cm11 Outreach Home Room NewsInformation Current HABFES OctoberEvan Racah861MayArtQuestions forFeature photo Tritium is

  13. Operating Experience Review of Tritium-in-Water Monitors

    SciTech Connect (OSTI)

    S. A. Bruyere; L. C. Cadwallader

    2011-09-01

    Monitoring tritium facility and fusion experiment effluent streams is an environmental safety requirement. This paper presents data on the operating experience of a solid scintillant monitor for tritium in effluent water. Operating experiences were used to calculate an average monitor failure rate of 4E-05/hour for failure to function. Maintenance experiences were examined to find the active repair time for this type of monitor, which varied from 22 minutes for filter replacement to 11 days of downtime while waiting for spare parts to arrive on site. These data support planning for monitor use; the number of monitors needed, allocating technician time for maintenance, inventories of spare parts, and other issues.

  14. Tritium Production from Palladium Alloys

    SciTech Connect (OSTI)

    Claytor, T.N.; Schwab, M.J.; Thoma, D.J.; Teter, D.F.; Tuggle, D.G.

    1998-04-19

    A number of palladium alloys have been loaded with deuterium or hydrogen under low energy bombardment in a system that allows the continuous measurement of tritium. Long run times (up to 200 h) result in an integration of the tritium and this, coupled with the high intrinsic sensitivity of the system ({approximately}0.1 nCi/l), enables the significance of the tritium measurement to be many sigma (>10). We will show the difference in tritium generation rates between batches of palladium alloys (Rh, Co, Cu, Cr, Ni, Be, B, Li, Hf, Hg and Fe) of various concentrations to illustrate that tritium generation rate is dependent on alloy type as well as within a specific alloy, dependent on concentration.

  15. RESULTS OF THE EXTRACTION-SCRUB-STRIP TESTING USING AN IMPROVED SOLVENT FORMULATION AND SALT WASTE PROCESSING FACILITY SIMULATED WASTE

    SciTech Connect (OSTI)

    Peters, T.; Washington, A.; Fink, S.

    2012-01-09

    The Office of Waste Processing, within the Office of Technology Innovation and Development, is funding the development of an enhanced solvent - also known as the next generation solvent (NGS) - for deployment at the Savannah River Site to remove cesium from High Level Waste. The technical effort is a collaborative effort between Oak Ridge National Laboratory (ORNL) and Savannah River National Laboratory (SRNL). As part of the program, the Savannah River National Laboratory (SRNL) has performed a number of Extraction-Scrub-Strip (ESS) tests. These batch contact tests serve as first indicators of the cesium mass transfer solvent performance with actual or simulated waste. The test detailed in this report used simulated Tank 49H material, with the addition of extra potassium. The potassium was added at 1677 mg/L, the maximum projected (i.e., a worst case feed scenario) value for the Salt Waste Processing Facility (SWPF). The results of the test gave favorable results given that the potassium concentration was elevated (1677 mg/L compared to the current 513 mg/L). The cesium distribution value, DCs, for extraction was 57.1. As a comparison, a typical D{sub Cs} in an ESS test, using the baseline solvent formulation and the typical waste feed, is {approx}15. The Modular Caustic Side Solvent Extraction Unit (MCU) uses the Caustic-Side Solvent Extraction (CSSX) process to remove cesium (Cs) from alkaline waste. This process involves the use of an organic extractant, BoBCalixC6, in an organic matrix to selectively remove cesium from the caustic waste. The organic solvent mixture flows counter-current to the caustic aqueous waste stream within centrifugal contactors. After extracting the cesium, the loaded solvent is stripped of cesium by contact with dilute nitric acid and the cesium concentrate is transferred to the Defense Waste Processing Facility (DWPF), while the organic solvent is cleaned and recycled for further use. The Salt Waste Processing Facility (SWPF), under construction, will use the same process chemistry. The Office of Waste Processing (EM-31) expressed an interest in investigating the further optimization of the organic solvent by replacing the BoBCalixC6 extractant with a more efficient extractant. This replacement should yield dividends in improving cesium removal from the caustic waste stream, and in the rate at which the caustic waste can be processed. To that end, EM-31 provided funding for both the Savannah River National Laboratory (SRNL) and the Oak Ridge National Laboratory (ORNL). SRNL wrote a Task Technical Quality and Assurance Plan for this work. As part of the envisioned testing regime, it was decided to perform an ESS test using a simulated waste that simulated a typical envisioned SWPF feed, but with added potassium to make the waste more challenging. Potassium interferes in the cesium removal, and its concentration is limited in the feed to <1950 mg/L. The feed to MCU has typically contained <500 mg/L of potassium.

  16. Tritium Behavior in Lead Lithium Eutectic (LLE) at Low Tritium...

    Office of Environmental Management (EM)

    governing these phenomenon are non-linear and a Newton solver is used to converge the equation set each time-step M.Shimada | Tritium Focus Group meeting | SRNL, SC | April 25,...

  17. Differential atmospheric tritium sampler

    DOE Patents [OSTI]

    Griesbach, O.A.; Stencel, J.R.

    1987-10-02

    An atmospheric tritium sampler is provided which uses a carrier gas comprised of hydrogen gas and a diluting gas, mixed in a nonexplosive concentration. Sample air and carrier gas are drawn into and mixed in a manifold. A regulator meters the carrier gas flow to the manifold. The air sample/carrier gas mixture is pulled through a first moisture trap which adsorbs water from the air sample. The moisture then passes through a combustion chamber where hydrogen gas in the form of H/sub 2/ or HT is combusted into water. The manufactured water is transported by the air stream to a second moisture trap where it is adsorbed. The air is then discharged back into the atmosphere by means of a pump.

  18. Particulate Generation in a Tritium System | Department of Energy

    Office of Environmental Management (EM)

    Particulate Generation in a Tritium System Particulate Generation in a Tritium System Presentation from the 33rd Tritium Focus Group Meeting held in Aiken, South Carolina on April...

  19. Memorandum for Tritium Focus Group Members from Bill Weaver

    Broader source: Energy.gov [DOE]

    Official Position of the Tritium Focus Group on Hazard Category 2 and 3 Threshold Values for Tritium.

  20. Tritium emissions from 200 East Area Double-Shell Tanks

    SciTech Connect (OSTI)

    Bachand, D.D.

    1994-11-28

    This document evaluates the need for tritium sampling of the emissions from the 200 East Area Double Shell Tanks based on the requirements of {open_quotes}National Emission Standards for Hazardous Air Pollutants{close_quotes} (NESHAP). The NESHAP requirements are specified in 40 Code of Federal Regulation (CFR), Part 61, Subpart H; {open_quotes}National Emission Standards for Emissions of Radionuclides Other than Radon from Department of Energy Facilities{close_quotes}.

  1. Development of a Novel Contamination Resistant Ionchamber for Process Tritium Measurement and use in the JET First Trace Tritium Experiment

    E-Print Network [OSTI]

    Development of a Novel Contamination Resistant Ionchamber for Process Tritium Measurement and use in the JET First Trace Tritium Experiment

  2. Summary of Topic1 Fusion Power Extraction

    E-Print Network [OSTI]

    California at Los Angeles, University of

    production · EU: HCLL, HCPB, WCLL, DCLL · US: DCLL (HCCB) · KO: HCCB (LL) CH: HCLL, DCLL, HCCB · IN: LCCB&D and facilities strongly overlap RAFM Steel PbLi Breeder Helium Cooled Ceramic Breeder Beryllium Helium Cooled PbLi Breeder Water Cooled Helium Coolant PbLi/ Tritium CB/ Tritium Two classes of concepts: Liquid Breeder

  3. CHARTER OF THE TRITIUM FOCUS GROUP (TFG)

    Office of Environmental Management (EM)

    OF THE TRITIUM FOCUS GROUP (TFG) APRIL 2013 PURPOSE - The purpose of the TFG, a Standing DOE Working Group, is to promote cost-effective improvements in tritium safety,...

  4. Crediting Tritium Deposition in Accident Analysis

    SciTech Connect (OSTI)

    Murphy, C.E. Jr.

    2001-06-20

    This paper describes the major aspects of tritium dispersion phenomenology, summarizes deposition attributes of the computer models used in the DOE Complex for tritium dispersion, and recommends an approach to account for deposition in accident analysis.

  5. Overview of Recent Tritium Experiments in TPE

    SciTech Connect (OSTI)

    Masashi Shimada; T. Otsuka; R. J. Pawelko; P. Calderoni; J. P. Sharpe

    2010-10-01

    Tritium retention in plasma-facing components influences the design, operation, and lifetime of fusion devices such as ITER. Most of the retention studies were carried out with the use of either hydrogen or deuterium. Tritium Plasma Experiment is a unique linear plasma device that can handle radioactive fusion fuel of tritium, toxic material of beryllium, and neutron-irradiated material. A tritium depth profiling method up to mm range was developed using a tritium imaging plate and a diamond wire saw. A series of tritium experiments (T2/D2 ratio: 0.2 and 0.5 %) was performed to investigate tritium depth profiling in bulk tungsten, and the results shows that tritium is migrated into bulk tungsten up to mm range.

  6. Tritium Separation at Cernavoda Nuclear – Romania

    Broader source: Energy.gov [DOE]

    Presentation from the 35th Tritium Focus Group Meeting held in Princeton New Jersey on May 05-07, 2015.

  7. Tritium Gas Processing for Magnetic Fusion

    Broader source: Energy.gov [DOE]

    Presentation from the 33rd Tritium Focus Group Meeting held in Aiken, South Carolina on April 22-24, 2014.

  8. Secure Wireless Tritium Air Monitoring Cart Development

    Broader source: Energy.gov [DOE]

    Presentation from the 33rd Tritium Focus Group Meeting held in Aiken, South Carolina on April 22-24, 2014.

  9. A Plan for Modularization of Tritium Components

    Office of Energy Efficiency and Renewable Energy (EERE)

    Presentation from the 33rd Tritium Focus Group Meeting held in Aiken, South Carolina on April 22-24, 2014.

  10. Tritium High Vacuum Pump Test Plan

    Broader source: Energy.gov [DOE]

    Presentation from the 33rd Tritium Focus Group Meeting held in Aiken, South Carolina on April 22-24, 2014.

  11. Tritium Sessions At The 2012 ANS Meeting

    Broader source: Energy.gov [DOE]

    Presentation from the 32nd Tritium Focus Group Meeting held in Germantown, Maryland on April 23-25, 2013.

  12. Monitoring of Tritium release at PTC

    Broader source: Energy.gov [DOE]

    Presentation from the 34th Tritium Focus Group Meeting held in Idaho Falls, Idaho on September 23-25, 2014.

  13. Chalk River Tritium Activities: Select Topics

    Broader source: Energy.gov [DOE]

    Presentation from the 35th Tritium Focus Group Meeting held in Princeton, New Jersey on May 05-07, 2015.

  14. Tritium Leak Detection: Strategies and Applications

    Broader source: Energy.gov [DOE]

    Presentation from the 33rd Tritium Focus Group Meeting held in Aiken, South Carolina on April 22-24, 2014.

  15. Overview of tritium activity in Japan

    Broader source: Energy.gov [DOE]

    Presentation from the 34th Tritium Focus Group Meeting held in Idaho Falls, Idaho on September 23-25, 2014.

  16. Tritium APEX Interim Report November, 1999

    E-Print Network [OSTI]

    California at Los Angeles, University of

    to the accepted level of ~ 10 Ci/d, the tritium partial pressure facing the primary heat exchanger has to be lower the Flibe to a very low inventory is not an issue. The issue is to reduce the tritium partial pressure so, the allowable tritium partial pressure will still have to be lower than 10 -5 Pa. A vacuum disengager

  17. EFFECTS OF TRITIUM GAS EXPOSURE ON POLYMERS

    SciTech Connect (OSTI)

    Clark, E.; Fox, E.; Kane, M.; Staack, G.

    2011-01-07

    Effects of tritium gas exposure on various polymers have been studied over the last several years. Despite the deleterious effects of beta exposure on many material properties, structural polymers continued to be used in tritium systems. Improved understanding of the tritium effects will allow more resistant materials to be selected. Currently polymers find use mainly in tritium gas sealing applications (eg. valve stem tips, O-rings). Future uses being evaluated including polymeric based cracking of tritiated water, and polymer-based sensors of tritium.

  18. In-vessel tritium retention and removal in ITER

    SciTech Connect (OSTI)

    Federici, G.; Anderl, R.A.; Andrew, P.

    1998-06-01

    The International Thermonuclear Experimental Reactor (ITER) is envisioned to be the next major step in the world`s fusion program from the present generation of tokamaks and is designed to study fusion plasmas with a reactor relevant range of plasma parameters. During normal operation, it is expected that a fraction of the unburned tritium, that is used to routinely fuel the discharge, will be retained together with deuterium on the surfaces and in the bulk of the plasma facing materials (PFMs) surrounding the core and divertor plasma. The understanding of he basic retention mechanisms (physical and chemical) involved and their dependence upon plasma parameters and other relevant operation conditions is necessary for the accurate prediction of the amount of tritium retained at any given time in the ITER torus. Accurate estimates are essential to assess the radiological hazards associated with routine operation and with potential accident scenarios which may lead to mobilization of tritium that is not tenaciously held. Estimates are needed to establish the detritiation requirements for coolant water, to determine the plasma fueling and tritium supply requirements, and to establish the needed frequency and the procedures for tritium recovery and clean-up. The organization of this paper is as follows. Section 2 provides an overview of the design and operating conditions of the main components which define the plasma boundary of ITER. Section 3 reviews the erosion database and the results of recent relevant experiments conducted both in laboratory facilities and in tokamaks. These data provide the experimental basis and serve as an important benchmark for both model development (discussed in Section 4) and calculations (discussed in Section 5) that are required to predict tritium inventory build-up in ITER. Section 6 emphasizes the need to develop and test methods to remove the tritium from the codeposited C-based films and reviews the status and the prospects of the most attractive techniques. Section 7 identifies the unresolved issues and provides some recommendations on potential R and D avenues for their resolution. Finally, a summary is provided in Section 8.

  19. Fabrication, properties, and tritium recovery from solid breeder materials

    SciTech Connect (OSTI)

    Johnson, C.E. (Argonne National Lab., IL (USA)); Kondo, T. (Japan Atomic Energy Research Inst., Tokyo (Japan)); Roux, N. (CEA Centre d'Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)); Tanaka, S. (Tokyo Univ. (Japan)); Vollath, D. (Kernforschungszentrum Karlsruhe GmbH (Germany, F.R.))

    1991-01-01

    The breeding blanket is a key component of the fusion reactor because it directly involves tritium breeding and energy extraction, both of which are critical to development of fusion power. The lithium ceramics continue to show promise as candidate breeder materials. This promise was recognized by the International Thermonuclear Experimental Reactor (ITER) design team in its selection of ceramics as the first option for the ITER breeder material. Blanket design studies have indicated properties in the candidate materials data base that need further investigation. Current studies are focusing on tritium release behavior at high burnup, changes in thermophysical properties with burnup, compatibility between the ceramic breeder and beryllium multiplier, and phase changes with burnup. Laboratory and in-reactor tests, some as part of an international collaboration for development of ceramic breeder materials, are underway. 133 refs., 1 fig.

  20. EFFECTS OF TRITIUM EXPOSURE ON UHMW-PE, PTFE, AND VESPEL

    SciTech Connect (OSTI)

    Clark, E; Kirk Shanahan, K

    2006-05-31

    Samples of three polymers, Ultra-High Molecular Weight Polyethylene (UHMW-PE), polytetrafluoroethylene (PTFE, also known as Teflon{reg_sign}), and Vespel{reg_sign} polyimide were exposed to 1 atmosphere of tritium gas at ambient temperature for varying times up to 2.3 years in closed containers. Sample mass and size measurements (to calculate density), spectra-colorimetry, dynamic mechanical analysis (DMA), and Fourier-transform infrared spectroscopy (FT-IR) were employed to characterize the effects of tritium exposure on these samples. Changes of the tritium exposure gas itself were characterized at the end of exposure by measuring total pressure and by mass spectroscopic analysis of the gas composition. None of the polymers exhibited significant changes of density. The color of initially white UHMW-PE and PTFE dramatically darkened to the eye and the color also significantly changed as measured by colorimetry. The bulk of UHMW-PE darkened just like the external surfaces, however the fracture surface of PTFE appeared white compared to the PTFE external surfaces. The white interior could have been formed while the sample was breaking or could reflect the extra tritium dose at the surface directly from the gas. The dynamic mechanical response of UHMW-PE was typical of radiation effects on polymers- an initial stiffening (increased storage modulus) and reduction of viscous behavior after three months exposure, followed by lowering of the storage modulus after one year exposure and longer. The storage modulus of PTFE increased through about nine months tritium exposure, then the samples became too weak to handle or test using DMA. Characterization of Vespel{reg_sign} using DMA was problematic--sample-to-sample variations were significant and no systematic change with tritium exposure could be discerned. Isotopic exchange and incorporation of tritium into UHMW-PE (exchanging for protium) and into PTFE (exchanging for fluorine) was observed by FT-IR using an attenuated total reflectance method. No significant change in the Vespel{reg_sign} infrared spectrum was observed after three months exposure. Protium significantly pressurized the UHMW-PE containers during exposure to about nine atmospheres (the initial pressure was one atmosphere of tritium). This is consistent with the well-known production of hydrogen by irradiation of polyethylene by ionizing radiation. The total pressure in the PTFE containers decreased, and a mass balance reveals that the observed decrease is consistent with the formation of small amounts of {sup 3}HF, which is condensed at ambient temperature. No significant change of pressure occurred in the Vespel{reg_sign} containers; however the composition of the gas became about 50% protium, showing that Vespel{reg_sign} interacted with the tritium gas atmosphere to some degree. The relative resistance to degradation from tritium exposure is least for PTFE, more for UHMW-PE, and the most for Vespel{reg_sign}, which is consistent with the known relative resistance of these polymers to gamma irradiation. This qualitatively agrees with the concept of equivalent effects for equivalent absorbed doses of radiation damage of polymers. Some of the changes of different polymers are qualitatively similar; however each polymer exhibited unique property changes when exposed to tritium. Information from this study that can be applied to a tritium facility is: (1) the relative resistance to tritium degradation of the three polymers studied is the same as the relative resistance to gamma irradiation in air (so relative rankings of polymer resistance to ionizing radiation can be used as a relative ranking for assessing tritium compatibility and polymer selection); and (2) all three polymers changed the gas atmosphere during tritium exposure--UHMW-PE and Vespel{reg_sign} exposed to tritium formed H{sub 2} gas (UHMW-PE much more so), and PTFE exposed to tritium formed {sup 3}HF. This observation of forming {sup 3}HF supports the general concept of minimizing chlorofluorocarbon polymers in tritium systems.

  1. DEPLOYMENT OF THE BULK TRITIUM SHIPPING PACKAGE

    SciTech Connect (OSTI)

    Blanton, P.

    2013-10-10

    A new Bulk Tritium Shipping Package (BTSP) was designed by the Savannah River National Laboratory to be a replacement for a package that has been used to ship tritium in a variety of content configurations and forms since the early 1970s. The BTSP was certified by the National Nuclear Safety Administration in 2011 for shipments of up to 150 grams of Tritium. Thirty packages were procured and are being delivered to various DOE sites for operational use. This paper summarizes the design features of the BTSP, as well as associated engineered material improvements. Fabrication challenges encountered during production are discussed as well as fielding requirements. Current approved tritium content forms (gas and tritium hydrides), are reviewed, as well as, a new content, tritium contaminated water on molecular sieves. Issues associated with gas generation will also be discussed.

  2. Tritium Behavior in Lead Lithium Eutectic (LLE) at Low Tritium Partial Pressure

    Office of Energy Efficiency and Renewable Energy (EERE)

    Presentation from the 33rd Tritium Focus Group Meeting held in Aiken, South Carolina on April 22-24, 2014.

  3. Tritium 2016 11TH International Conference on Tritium Science and Technology

    Office of Energy Efficiency and Renewable Energy (EERE)

    Presentation from the 33rd Tritium Focus Group Meeting held in Aiken, South Carolina on April 22-24, 2014.

  4. Is Tritium Over-Regulated, Part 2 Should The TFG Support Higher Tritium Threshold Values?

    Office of Energy Efficiency and Renewable Energy (EERE)

    Presentation from the 33rd Tritium Focus Group Meeting held in Aiken, South Carolina on April 22-24, 2014.

  5. Application of 3D Code IBSimu for Designing an H{sup -}/D{sup -} Extraction System for the Texas A and M Facility Upgrade

    SciTech Connect (OSTI)

    Kalvas, T.; Tarvainen, O.; Aerje, J. [Department of Physics, University of Jyvaeskylae, Jyvaeskylae, 40500 (Finland); Clark, H.; Brinkley, J. [Texas A and M University, Cyclotron Institute, College Station, TX 77843 (United States)

    2011-09-26

    A three dimensional ion optical code IBSimu is being developed at the University of Jyvaeskylae. So far the plasma modelling of the code has been restricted to positive ion extraction systems, but now a negative ion plasma extraction model has been added. The plasma model has been successfully validated with simulations of the Spallation Neutron Source (SNS) ion source extraction both in cylindrical symmetry and in full 3D, also modelling electron beam dumping and ion beam tilt. A filament-driven multicusp ion source has been installed at the Texas A and M University Cyclotron Institute for production of H{sup -} and D{sup -} beams as a part of the facility upgrade. The light ion beams, produced by the ion source, are accelerated with the K150 cyclotron for production and reacceleration of rare isotopes. The extraction system for the ion source was designed with IBSimu. The extraction features a water-cooled puller electrode with a permanent magnet dipole field for dumping the co-extracted electrons and a decelerating Einzel lens for adjusting the beam focusing for further beam transport. The ion source and the puller electrode are tilted at 4 degree angle with respect to the beam line. The extraction system can handle H{sup -} and D{sup -} beams with final beam energies from 5 keV to 15 keV using the same geometry, only adjusting the electrode voltages. So far, 24 {mu}A of H{sup -} and 15 {mu}A of D{sup -} have been extracted from the cyclotron.

  6. Tritium radioluminescent devices, Health and Safety Manual

    SciTech Connect (OSTI)

    Traub, R.J.; Jensen, G.A.

    1995-06-01

    This document consolidates available information on the properties of tritium, including its environmental chemistry, its health physics, and safe practices in using tritium-activated RL lighting. It also summarizes relevant government regulations on RL lighting. Chapters are divided into a single-column part, which provides an overview of the topic for readers simply requiring guidance on the safety of tritium RL lighting, and a dual-column part for readers requiring more technical and detailed information.

  7. Recovery of tritium from tritiated molecules

    DOE Patents [OSTI]

    Swansiger, W.A.

    1984-10-17

    This invention relates to the recovery of tritium from various tritiated molecules by reaction with uranium. More particularly, the invention relates to the recovery of tritium from tritiated molecules by reaction with uranium wherein the reaction is conducted in a reactor which permits the reaction to occur as a moving front reaction from the point where the tritium enters the reactor charged with uranium down the reactor until the uranium is exhausted.

  8. Reclassification of the Tritium Research Laboratory

    SciTech Connect (OSTI)

    Johnson, A.J.

    1997-01-01

    This document is a collection of the required actions that were taken to reclassify Building 968, the Tritium Research Laboratory, at Sandia National Laboratories/California.

  9. Methods for Post Irradiation Examination of Tritium Producing...

    Office of Environmental Management (EM)

    Methods for Post Irradiation Examination of Tritium Producing Burnable Absorber Rods Methods for Post Irradiation Examination of Tritium Producing Burnable Absorber Rods...

  10. The Princeton Tritium Observatory for Light, Early Universe,...

    Office of Environmental Management (EM)

    The Princeton Tritium Observatory for Light, Early Universe, Massive Neutrino Yield (PTOLEMY) The Princeton Tritium Observatory for Light, Early Universe, Massive Neutrino Yield...

  11. Tritium research laboratory cleanup and transition project final report

    SciTech Connect (OSTI)

    Johnson, A.J.

    1997-02-01

    This Tritium Research Laboratory Cleanup and Transition Project Final Report provides a high-level summary of this project`s multidimensional accomplishments. Throughout this report references are provided for in-depth information concerning the various topical areas. Project related records also offer solutions to many of the technical and or administrative challenges that such a cleanup effort requires. These documents and the experience obtained during this effort are valuable resources to the DOE, which has more than 1200 other process contaminated facilities awaiting cleanup and reapplication or demolition.

  12. TRITIUM AGING EFFECTS ON THE FRACTURE TOUGHNESS PROPERTIES OF STAINLESS STEEL BASE METAL AND WELDS

    SciTech Connect (OSTI)

    Morgan, M.

    2009-07-30

    Tritium reservoirs are constructed from welded stainless steel forgings. While these steels are highly resistant to the embrittling effects of hydrogen isotopes and helium from tritium decay; they are not immune. Tritium embrittlement is an enhanced form of hydrogen embrittlement because of the presence of helium-3 from tritium decay which nucleates as nanometer-sized bubbles on dislocations, grain boundaries, and other microstructural defects. Steels with decay helium bubble microstructures are hardened and less able to deform plastically and become more susceptible to embrittlement by hydrogen and its isotopes. Ductility, elongation-to-failure, and fracture toughness are reduced by exposures to tritium and the reductions increase with time as helium-3 builds into the material from tritium permeation and radioactive decay. Material and forging specifications have been developed for optimal material compatibility with tritium. These specifications cover composition, mechanical properties, and select microstructural characteristics like grain size, flow-line orientation, inclusion content, and ferrite distribution. For many years, the forming process of choice for reservoir manufacturing was high-energy-rate forging (HERF), principally because the DOE forging facility owned only HERF hammers. Today, some reservoir forgings are being made that use a conventional, more common process known as press forging (PF or CF). One of the chief differences between the two forging processes is strain rate: Conventional hydraulic or mechanical forging presses deform the metal at 4-8 ft/s, about ten-fold slower than the HERF process. The material specifications continue to provide successful stockpile performance by ensuring that the two forging processes produce similar reservoir microstructures. While long-term life storage tests have demonstrated the general tritium compatibility of tritium reservoirs, fracture-toughness properties of both conventionally forged and high-energy-rate forged are needed for designing and establishing longer tritium-reservoir lifetimes, ranking materials, and, potentially, for qualifying new forging vendors or processes. Measurements on the effects of tritium and decay helium on the fracture toughness properties of CF stainless steels having similar composition, grain size, and mechanical properties to previously studied HERF steels are needed and have not been conducted until now. The compatibility of stainless steel welds with tritium represents another concern for long-term reservoir performance. Weldments have not been well-characterized with respect to tritium embrittlement, although a recent study was completed on the effect of tritium and decay helium on the fracture toughness properties of Type 304L weldments. This study expands the characterization of weldments through measurements of tritium and decay helium effects on the fracture toughness properties of Type 21-6-9 stainless steel. The purpose of this study was to measure and compare the fracture toughness properties of Type 21-6-9 stainless steel for conventional forgings and weldments in the non-charged, hydrogen-charged and tritium-charged-and-aged conditions.

  13. The Use of Subsurface Barriers to Support Treatment of Metals and Reduce the Flux of Tritium to Fourmile Branch at the Savannah River Site in South Carolina - 13358

    SciTech Connect (OSTI)

    Blount, Gerald; Thibault, Jeffrey; Wells, Leslie [Savannah River Nuclear Solutions LLC, 730-4B, Aiken, SC 29808 (United States)] [Savannah River Nuclear Solutions LLC, 730-4B, Aiken, SC 29808 (United States); Prater, Phillip [Department of Energy, Savannah River Site, Aiken, SC 29808 (United States)] [Department of Energy, Savannah River Site, Aiken, SC 29808 (United States)

    2013-07-01

    The Savannah River Site (SRS) produced tritium, plutonium, and special nuclear materials for national defense, medicine, and the space programs. Acidic groundwater plumes containing metals, metallic radionuclides, non-metallic radionuclides and tritium sourced from the F and H Area Seepage Basins have impacted the surface water of Fourmile Branch on SRS. Tritium releases from Fourmile Branch have impacted the water quality within areas of the Savannah River adjacent to the SRS, and this circumstance has been an ongoing regulatory concern. The F and H Area Seepage Basins operated until 1988 for the disposition of deionized acidic waste water from the F and H Separations Facilities. The waste water contained dilute nitric acid and low concentrations of non-radioactive metals, and radionuclides, with the major isotopes being Cs-137, Sr-90, U-235, U-238, Pu-239, Tc-99, I-129, and tritium. The tritium concentration in the waste water was relatively elevated because there is not a practicable removal method in water. The acid content of the waste water during the operational period of the basins was equal to 12 billion liters of nitric acid. The seepage basins were closed in 1988 and backfilled and capped by 1991. The plumes associated with the F and H basins cover an area of nearly 2.4 square kilometers (600 acres) and discharge along ?2,600 meters of Fourmile Branch. The acidic nature of the plumes and their overall discharge extent along the branch represent a large challenge with respect to reducing contaminant flux to Fourmile Branch. The introduction of nitric acid into the groundwater over a long time effectively reduced the retardation of metal migration from the basins to the groundwater and in the groundwater to Fourmile Branch, because most negatively charged surfaces on the aquifer materials were filled with hydrogen ion. Two large pump and treat systems were constructed in 1997 and operated until 2003 in an attempt to capture and control the releases to Fourmile Branch. The operating cost, including waste disposal, for the two systems was ?$1.3 M/month. Both systems employed reinjection of tritiated water up gradient of the extraction, and produced large quantities of waste from non-tritium isotopes and metals removal prior to reinjection. Both systems were determined to be ineffective and potentially detrimental with respect to limiting the flux of contaminants to Fourmile Branch. After it became apparent that there was very little benefit to continued operation of the systems, and the staggering cost of operations was recognized by the SRS and regulators, a new remedy was developed. The new system uses vertical subsurface barriers to redirect groundwater flow to limit the transport of contaminants to the stream. The barriers were constructed of acid resistant grout using deep soil mixing techniques. The grout mixture used low swelling clay, fly ash, and sodium hydroxide to form a pozzolana material with low permeability and low strength. The SRS and regulators agreed to a series of remedial goals, with the first goal to reduce tritium flux to the stream by 70% and bring constituents other than tritium to groundwater protection standards. (authors)

  14. DOE handbook: Tritium handling and safe storage

    SciTech Connect (OSTI)

    1999-03-01

    The DOE Handbook was developed as an educational supplement and reference for operations and maintenance personnel. Most of the tritium publications are written from a radiological protection perspective. This handbook provides more extensive guidance and advice on the null range of tritium operations. This handbook can be used by personnel involved in the full range of tritium handling from receipt to ultimate disposal. Compliance issues are addressed at each stage of handling. This handbook can also be used as a reference for those individuals involved in real time determination of bounding doses resulting from inadvertent tritium releases. This handbook provides useful information for establishing processes and procedures for the receipt, storage, assay, handling, packaging, and shipping of tritium and tritiated wastes. It includes discussions and advice on compliance-based issues and adds insight to those areas that currently possess unclear DOE guidance.

  15. Tritium issues in commercial pressurized water reactors

    SciTech Connect (OSTI)

    Jones, G. [Constellation Energy Group, R.E. Ginna Nuclear Power Plant, Ontario, NY (United States)

    2008-07-15

    Tritium has become an important radionuclide in commercial Pressurized Water Reactors because of its mobility and tendency to concentrate in plant systems as tritiated water during the recycling of reactor coolant. Small quantities of tritium are released in routine regulated effluents as liquid water and as water vapor. Tritium has become a focus of attention at commercial nuclear power plants in recent years due to inadvertent, low-level, chronic releases arising from routine maintenance operations and from component failures. Tritium has been observed in groundwater in the vicinity of stations. The nuclear industry has undertaken strong proactive corrective measures to prevent recurrence, and continues to eliminate emission sources through its singular focus on public safety and environmental stewardship. This paper will discuss: production mechanisms for tritium, transport mechanisms from the reactor through plant, systems to the environment, examples of routine effluent releases, offsite doses, basic groundwater transport and geological issues, and recent nuclear industry environmental and legal ramifications. (authors)

  16. Thermal Release of 3He from Tritium Aged LaNi4.25Al0.75 Hydride

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Staack, Gregory C.; Crowder, Mark L.; Klein, James E.

    2015-02-01

    Recently, the demand for He-3 has increased dramatically due to widespread use in nuclear nonproliferation, cryogenic, and medical applications. Essentially all of the world’s supply of He-3 is created by the radiolytic decay of tritium. The Savannah River Site Tritium Facilities (SRS-TF) utilizes LANA.75 in the tritium process to store hydrogen isotopes. The vast majority of He-3 “born” from tritium stored in LANA.75 is trapped in the hydride metal matrix. The SRS-TF has multiple LANA.75 tritium storage beds that have been retired from service with significant quantities of He-3 trapped in the metal. To support He-3 recovery, the Savannah Rivermore »National Laboratory (SRNL) conducted thermogravimetric analysis coupled with mass spectrometry (TGA-MS) on a tritium aged LANA.75 sample. TGA-MS testing was performed in an argon environment. Prior to testing, the sample was isotopically exchanged with deuterium to reduce residual tritium and passivated with air to alleviate pyrophoric concerns associated with handling the material outside of an inert glovebox. Analyses indicated that gas release from this sample was bimodal, with peaks near 220 and 490°C. The first peak consisted of both He-3 and residual hydrogen isotopes, the second was primarily He-3. The bulk of the gas was released by 600 °C« less

  17. Tritium monitor with improved gamma-ray discrimination

    DOE Patents [OSTI]

    Cox, S.A.; Bennett, E.F.; Yule, T.J.

    1982-10-21

    Apparatus and method are presented for selective measurement of tritium oxide in an environment which may include other radioactive components and gamma radiation, the measurement including the selective separation of tritium oxide from a sample gas through a membrane into a counting gas, the generation of electrical pulses individually representative by rise times of tritium oxide and other radioactivity in the counting gas, separation of the pulses by rise times, and counting of those pulses representative of tritium oxide. The invention further includes the separate measurement of any tritium in the sample gas by oxidizing the tritium to tritium oxide and carrying out a second separation and analysis procedure as described above.

  18. Tritium monitor with improved gamma-ray discrimination

    DOE Patents [OSTI]

    Cox, Samson A. (Downers Grove, IL); Bennett, Edgar F. (Downers Grove, IL); Yule, Thomas J. (West Chicago, IL)

    1985-01-01

    Apparatus and method for selective measurement of tritium oxide in an environment which may include other radioactive components and gamma radiation, the measurement including the selective separation of tritium oxide from a sample gas through a membrane into a counting gas, the generation of electrical pulses individually representative by rise times of tritium oxide and other radioactivity in the counting gas, separation of the pulses by rise times, and counting of those pulses representative of tritium oxide. The invention further includes the separate measurement of any tritium in the sample gas by oxidizing the tritium to tritium oxide and carrying out a second separation and analysis procedure as described above.

  19. Evaluation of selected ex-reactor accidents related to the tritium and medical isotope production mission at the FFTF

    SciTech Connect (OSTI)

    Himes, D.A.

    1997-11-17

    The Fast Flux Test Facility (FFTF) has been proposed as a production facility for tritium and medical isotopes. A range of postulated accidents related to ex-reactor irradiated fuel and target handling were identified and evaluated using new source terms for the higher fuel enrichment and for the tritium and medical isotope targets. In addition, two in-containment sodium spill accidents were re-evaluated to estimate effects of increased fuel enrichment and the presence of the Rapid Retrieval System. Radiological and toxicological consequences of the analyzed accidents were found to be well within applicable risk guidelines.

  20. THERMAL ENHANCEMENT CARTRIDGE HEATER MODIFIED TECH MOD TRITIUM HYDRIDE BED DEVELOPMENT PART I DESIGN AND FABRICATION

    SciTech Connect (OSTI)

    Klein, J.; Estochen, E.

    2014-03-06

    The Savannah River Site (SRS) tritium facilities have used 1{sup st} generation (Gen1) LaNi{sub 4.25}Al{sub 0.75} (LANA0.75) metal hydride storage beds for tritium absorption, storage, and desorption. The Gen1 design utilizes hot and cold nitrogen supplies to thermally cycle these beds. Second and 3{sup rd} generation (Gen2 and Gen3) storage bed designs include heat conducting foam and divider plates to spatially fix the hydride within the bed. For thermal cycling, the Gen2 and Gen 3 beds utilize internal electric heaters and glovebox atmosphere flow over the bed inside the bed external jacket for cooling. The currently installed Gen1 beds require replacement due to tritium aging effects on the LANA0.75 material, and cannot be replaced with Gen2 or Gen3 beds due to different designs of these beds. At the end of service life, Gen1 bed desorption efficiencies are limited by the upper temperature of hot nitrogen supply. To increase end-of-life desorption efficiency, the Gen1 bed design was modified, and a Thermal Enhancement Cartridge Heater Modified (TECH Mod) bed was developed. Internal electric cartridge heaters in the new design to improve end-of-life desorption, and also permit in-bed tritium accountability (IBA) calibration measurements to be made without the use of process tritium. Additional enhancements implemented into the TECH Mod design are also discussed.

  1. Scale-Up of Palladium Powder Production Process for Use in the Tritium Facility at Westinghouse, Savannah River, SC/Summary of FY99-FY01 Results for the Preparation of Palladium Using the Sandia/LANL Process

    SciTech Connect (OSTI)

    David P. Baldwin; Daniel S. Zamzow; R. Dennis Vigil; Jesse T. Pikturna

    2001-08-24

    Palladium used at Savannah River (SR) for process tritium storage is currently obtained from a commercial source. In order to understand the processes involved in preparing this material, SR is supporting investigations into the chemical reactions used to synthesize this material. The material specifications are shown in Table 1. An improved understanding of the chemical processes should help to guarantee a continued reliable source of Pd in the future. As part of this evaluation, a work-for-others contract between Westinghouse Savannah River Company and Ames Laboratory (AL) was initiated. During FY98, the process for producing Pd powder developed in 1986 by Dan Grove of Mound Applied Technologies, USDOE (the Mound muddy water process) was studied to understand the processing conditions that lead to changes in morphology in the final product. During FY99 and FY00, the process for producing Pd powder that has been used previously at Sandia and Los Alamos National Laboratories (the Sandia/LANL process) was studied to understand the processing conditions that lead to changes in the morphology of the final Pd product. During FY01, scale-up of the process to batch sizes greater than 600 grams of Pd using a 20-gallon Pfaudler reactor was conducted by the Iowa State University (ISU) Chemical Engineering Department. This report summarizes the results of FY99-FY01 Pd processing work done at AL and ISU using the Sandia/LANL process. In the Sandia/LANL process, Pd is dissolved in a mixture of nitric and hydrochloric acids. A number of chemical processing steps are performed to yield an intermediate species, diamminedichloropalladium (Pd(NH{sub 3}){sub 2}Cl{sub 2}, or DADC-Pd), which is isolated. In the final step of the process, the Pd(NH{sub 3}){sub 2}Cl{sub 2} intermediate is subsequently redissolved, and Pd is precipitated by the addition of a reducing agent (RA) mixture of formic acid and sodium formate. It is at this point that the morphology of the Pd product is determined. During FY99 and FY00, a study of how the characteristics of the Pd are affected by changes in processing conditions including the RA/Pd molar ratio, Pd concentration, mole fraction of formic acid (mf-FA) in the RA solution, reaction temperature, and mixing was performed. These parameters all had significant effects on the resulting values of the tap density (TD), BET surface area (SA), and Microtrac particle size (PS) distribution for the Pd samples. These effects were statistically modeled and fit in order to determine ranges of predicted experimental conditions that resulted in material that meets the requirements for the Pd powder to be used at SR. Although not statistically modeled, the method and rate of addition of the RA and the method and duration of stirring were shown to be significant factors affecting the product morphology. Instead of producing an additional statistical fit and due to the likely changes anticipated during scale-up of this processing procedure, these latter conditions were incorporated into a reproducible practical method of synthesis. Palladium powder that met the SR specifications for TD, BET SA, and Microtrac PS was reliably produced at batch sizes ranging from 25-100 grams. In FY01, scale-up of the Sandia/LANL process was investigated by the ISU Chemical Engineering Department for the production of 600-gram batches of Pd. Palladium that meets the SR specifications for TD, BET SA, and Microtrac PS has been produced using the Pfaudler reactor, and additional processing batches will be done during the remainder of FY01 to investigate the range of conditions that can be used to produce Pd powder within specifications. Palladium product samples were analyzed at AL and SR to determine TD and at SR to determine BET SA, Microtrac PS distribution, and Pd nodule size and morphology by scanning electron microscopy (SEM).

  2. Tritium containing polymers having a polymer backbone substantially void of tritium

    DOE Patents [OSTI]

    Jensen, G.A.; Nelson, D.A.; Molton, P.M.

    1992-03-31

    A radioluminescent light source comprises a solid mixture of a phosphorescent substance and a tritiated polymer. The solid mixture forms a solid mass having length, width, and thickness dimensions, and is capable of self-support. In one aspect of the invention, the phosphorescent substance comprises solid phosphor particles supported or surrounded within a solid matrix by a tritium containing polymer. The tritium containing polymer comprises a polymer backbone which is essentially void of tritium. 2 figs.

  3. Precision Shock Tuning on the National Ignition Facility

    E-Print Network [OSTI]

    Frenje, Johan A.

    Ignition implosions on the National Ignition Facility [ J.?D. Lindl et al. Phys. Plasmas 11 339 (2004)] are underway with the goal of compressing deuterium-tritium fuel to a sufficiently high areal density (?R) to sustain ...

  4. EIS-0161: Tritium Supply and Recycling

    Broader source: Energy.gov [DOE]

    This PEIS evaluates the potential environmental impacts of technology and siting alternatives for the production of tritium for national security purposes as well as the impacts of constructing a...

  5. Tritium evolution from various morphologies of palladium

    SciTech Connect (OSTI)

    Tuggle, D.G.; Claytor, T.N.; Taylor, S.F. |

    1994-04-01

    The authors have been able to extend the tritium production techniques to various novel morphologies of palladium. These include small solid wires of various diameters and a type of pressed powder wire and a plasma cell. In most successful experiments, the amount of palladium required, for an equivalent tritium output, has been reduced by a factor of 100 over the older powder methods. In addition, they have observed rates of tritium production (>5 nCi/h) that far exceed most of the previous results. Unfortunately, the methods that they currently use to obtain the tritium are poorly understood and consequently there are numerous variables that need to be investigated before the new methods are as reliable and repeatable as the previous techniques. For instance, it seems that surface and/or bulk impurities play a major role in the successful generation of any tritium. In those samples with total impurity concentrations of >400 ppM essentially no tritium has been generated by the gas loading and electrical simulation methods.

  6. Method and apparatus for controlling accidental releases of tritium

    DOE Patents [OSTI]

    Galloway, T.R.

    1980-04-01

    An improvement is described in a tritium control system based on a catalytic oxidation reactor wherein accidental releases of tritium into room air are controlled by flooding the catalytic oxidation reactor with hydrogen when the tritium concentration in the room air exceeds a specified limit. The sudden flooding with hydrogen heats the catalyst to a high temperature within seconds, thereby greatly increasing the catalytic oxidation rate of tritium to tritiated water vapor. Thus, the catalyst is heated only when needed. In addition to the heating effect, the hydrogen flow also swamps the tritium and further reduces the tritium release. 1 fig.

  7. Method and apparatus for controlling accidental releases of tritium

    DOE Patents [OSTI]

    Galloway, Terry R. [Berkeley, CA

    1980-04-01

    An improvement in a tritium control system based on a catalytic oxidation reactor wherein accidental releases of tritium into room air are controlled by flooding the catalytic oxidation reactor with hydrogen when the tritium concentration in the room air exceeds a specified limit. The sudden flooding with hydrogen heats the catalyst to a high temperature within seconds, thereby greatly increasing the catalytic oxidation rate of tritium to tritiated water vapor. Thus, the catalyst is heated only when needed. In addition to the heating effect, the hydrogen flow also swamps the tritium and further reduces the tritium release.

  8. PPPL-3194 -Preprint: July 1996, UC-420, 421, 426 Tritium Recycling and Transport in TFTR Plasmas

    E-Print Network [OSTI]

    concentrations. Also, since it is difficult to reprocess tritium, it will be important to fuel the tritium

  9. EFFECTS OF TRITIUM GAS EXPOSURE ON ELECTRICALLY CONDUCTING POLYMERS

    SciTech Connect (OSTI)

    Kane, M.; Clark, E.; Lascola, R.

    2009-12-16

    Effects of beta (tritium) and gamma irradiation on the surface electrical conductivity of two types of conducting polymer films are documented to determine their potential use as a sensing and surveillance device for the tritium facility. It was shown that surface conductivity was significantly reduced by irradiation with both gamma and tritium gas. In order to compare the results from the two radiation sources, an approximate dose equivalence was calculated. The materials were also sensitive to small radiation doses (<10{sup 5} rad), showing that there is a measurable response to relatively small total doses of tritium gas. Spectroscopy was also used to confirm the mechanism by which this sensing device would operate in order to calibrate this sensor for potential use. It was determined that one material (polyaniline) was very sensitive to oxidation while the other material (PEDOT-PSS) was not. However, polyaniline provided the best response as a sensing material, and it is suggested that an oxygen-impermeable, radiation-transparent coating be applied to this material for future device prototype fabrication. A great deal of interest has developed in recent years in the area of conducting polymers due to the high levels of conductivity that can be achieved, some comparable to that of metals [Gerard 2002]. Additionally, the desirable physical and chemical properties of a polymer are retained and can be exploited for various applications, including light emitting diodes (LED), anti-static packaging, electronic coatings, and sensors. The electron transfer mechanism is generally accepted as one of electron 'hopping' through delocalized electrons in the conjugated backbone, although other mechanisms have been proposed based on the type of polymer and dopant [Inzelt 2000, Gerard 2002]. The conducting polymer polyaniline (PANi) is of particular interest because there are extensive studies on the modulation of the conductivity by changing either the oxidation state of the main backbone chain, or by protonation of the imine groups [de Acevedo, 1999]. There are several types of radiation sensors commercially available, including ionization chambers, geiger counters, proportional counters, scintillators and solid state detectors. Each type has advantages, although many of these sensors require expensive electronics for signal amplification, are large and bulky, have limited battery life or require expensive materials for fabrication. A radiation sensor constructed of a polymeric material could be flexible, light, and the geometry designed to suit the application. Very simple and inexpensive electronics would be necessary to measure the change in conductivity with exposure to radiation and provide an alarm system when a set change of conductivity occurs in the sensor that corresponds to a predetermined radiation dose having been absorbed by the polymer. The advantages of using a polymeric sensor of this type rather than those currently in use are the flexibility of sensor geometry and relatively low cost. It is anticipated that these sensors can be made small enough for glovebox applications or have the ability to monitor the air tritium levels in places where a traditional monitor cannot be placed. There have been a few studies on the changes in conductivity of polyaniline specifically for radiation detection [de Acevedo, 1999; Lima Pacheco, 2003], but there have been no reports on the effects of tritium (beta radiation) on conducting polymers, such as polyaniline or polythiophene. The direct implementation of conducting polymers as radiation sensor materials has not yet been commercialized due to differing responses with total dose, dose rate, etc. Some have reported a large increase in the surface conductivity with radiation dose while others report a marked decrease in conductive properties; these differing observations may reflect the competing mechanisms of chain scission and cross-linking. However, it is clear that the radiation dose effects on conducting polymers must be fully understood before these materials can be used

  10. Commercial Light Water Production of Tritium Update and Path...

    Office of Environmental Management (EM)

    Light Water Production of Tritium: Update and Path Forward Dave Senor April 23, 2013 Tritium Focus Group 1 PNNL-SA-94431 Background United States defense maintains a stockpile of...

  11. 32nd Tritium Focus Group Meeting, Cloverleaf Building, Germantown...

    Office of Environmental Management (EM)

    Building, Germantown MD, April 23-25, 2013 The Tritium Focus Group (TFG), is a long standing DOE Working Group, whose purpose is to promote cost-effective improvements in tritium...

  12. Charter of the Tritium Focus Group (TFG) | Department of Energy

    Office of Environmental Management (EM)

    Focus Group (TFG) Charter of the Tritium Focus Group (TFG) The purpose of the TFG, a Standing DOE Working Group, is to promote cost-effective improvements in tritium safety,...

  13. Accelerator driven production of tritium: target and blanket design 

    E-Print Network [OSTI]

    Ragusa, Jean Concetto

    1996-01-01

    Tritium is an essential component of thermonuclear weapons in the US arsenal. Unfortunately, tritium is a radioactive form of hydrogen, and one-half of the inventory disappears through radioactive decay every 12 years; therefore, it must...

  14. Tritium R&D at AECL Selected Topics

    Office of Environmental Management (EM)

    Tritium R&D at AECL Selected Topics Tritium Focus Group Meeting, Savannah River Site 2014 April 22-24 Hugh Boniface Chalk River Laboratories, Ontario, CANADA Outline of...

  15. Let's Compare Tritium Design Practices Across The DOE Complex

    Broader source: Energy.gov [DOE]

    Presentation from the 32nd Tritium Focus Group Meeting held in Germantown, Maryland on April 23-25, 2013.

  16. Tritium Transport within the TMIST-3 In-Reactor Experiment

    Broader source: Energy.gov [DOE]

    Presentation from the 35th Tritium Focus Group Meeting held in Princeton, New Jersey on May 05-07, 2015.

  17. Commercial Light Water Production of Tritium: Update and Path Forward

    Broader source: Energy.gov [DOE]

    Presentation from the 32nd Tritium Focus Group Meeting held in Germantown, Maryland on April 23-25, 2013.

  18. Tritium Plasma Experiment and Its Role in PHENIX Program

    Office of Energy Efficiency and Renewable Energy (EERE)

    Presentation from the 34th Tritium Focus Group Meeting held in Idaho Falls, Idaho on September 23-25, 2014.

  19. Overview of AECL’s Tritium Compatible Electrolyser Program

    Broader source: Energy.gov [DOE]

    Presentation from the 34th Tritium Focus Group Meeting held in Idaho Falls, Idaho on September 23-25, 2014.

  20. Tritium R&D at AECL Selected Topics

    Office of Energy Efficiency and Renewable Energy (EERE)

    Presentation from the 33rd Tritium Focus Group Meeting held in Aiken, South Carolina on April 22-24, 2014.

  1. Meeting Attendance- 36th Tritium Focus Group Meeting

    Broader source: Energy.gov [DOE]

    Attendees to the 36th Tritium Focus Group Meeting held in Los Alamos, New Mexico, November 03-05, 2015.

  2. Overview of Tritium Activities at the Laboratory for Laser Energetics

    Broader source: Energy.gov [DOE]

    Presentation from the 32nd Tritium Focus Group Meeting held in Germantown, Maryland on April 23-25, 2013.

  3. Meeting Attendance- 33rd Tritium Focus Group Meeting

    Broader source: Energy.gov [DOE]

    Attendees to the 33rd Tritium Focus Group Meeting, held in Aiken, South Carolina, April 22-24, 2014.

  4. Tritium Irrigation Facility & Automated Vadose Zone Monitoring System |

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of NaturalDukeWakefieldSulfateSciTechtail.Theory ofDidDevelopmentat LENA| Reaction Rates| UNCTandem Schedule|Savannah River

  5. Tritium Aging Effects in Palladium on Kieselguhr

    SciTech Connect (OSTI)

    Shanahan, K.L.; Holder, J.S.; Wermer, J.R.

    1998-10-01

    50 weight % Pd on kieselguhr (Pd/k) is used in hydrogen isotope separation processes at the Savannah River Site. Long term aging studies on this material were undertaken in June, 1992. P-c-T data showing the aging effect of tritium loading for long periods will be presented and discussed covering from June, 1992 to March, 1997. Lowering of plateau pressures and increasing indications of in homogeneities have been observed in both tritium and deuterium absorption isotherms at 0 C, and desorption isotherms at 80 and 120 C.

  6. Is Tritium over-regulated by DOE? Should the TFG support NA-1 SD G 1027 tritium values?

    Office of Energy Efficiency and Renewable Energy (EERE)

    Presentation from the 32nd Tritium Focus Group Meeting held in Germantown, Maryland on April 23-25, 2013.

  7. Relativistic Cyclotron Radiation Detection of Tritium Decay Electrons as a New Technique for Measuring the Neutrino Mass

    E-Print Network [OSTI]

    Benjamin Monreal; Joseph A. Formaggio

    2009-04-18

    The shape of the beta decay energy distribution is sensitive to the mass of the electron neutrino. Attempts to measure the endpoint shape of tritium decay have so far seen no distortion from the zero-mass form, thus placing an upper limit of m_nu_beta detect the coherent cyclotron radiation emitted by an energetic electron in a magnetic field. For mildly relativistic electrons, like those in tritium decay, the relativistic shift of the cyclotron frequency allows us to extract the electron energy from the emitted radiation. We present calculations for the energy resolution, noise limits, high-rate measurement capability, and systematic errors expected in such an experiment.

  8. EIS-0288-S1: Production of Tritium in a Commercial Light Water Reactor (CLWR) Tritium Readiness Supplemental Environmental Impact Statement

    Broader source: Energy.gov [DOE]

    This Supplemental EIS updates the environmental analyses in DOE’s 1999 EIS for the Production of Tritium in a Commercial Light Water Reactor (CLWR EIS). The CLWR EIS addressed the production of tritium in Tennessee Valley Authority reactors in Tennessee using tritium-producing burnable absorber rods.

  9. Small system for tritium accelerator mass spectrometry

    DOE Patents [OSTI]

    Roberts, M.L.; Davis, J.C.

    1993-02-23

    Apparatus for ionizing and accelerating a sample containing isotopes of hydrogen and detecting the ratios of hydrogen isotopes contained in the sample is disclosed. An ion source generates a substantially linear ion beam including ions of tritium from the sample. A radio-frequency quadrupole accelerator is directly coupled to and axially aligned with the source at an angle of substantially zero degrees. The accelerator accelerates species of the sample having different mass to different energy levels along the same axis as the ion beam. A spectrometer is used to detect the concentration of tritium ions in the sample. In one form of the invention, an energy loss spectrometer is used which includes a foil to block the passage of hydrogen, deuterium and [sup 3]He ions, and a surface barrier or scintillation detector to detect the concentration of tritium ions. In another form of the invention, a combined momentum/energy loss spectrometer is used which includes a magnet to separate the ion beams, with Faraday cups to measure the hydrogen and deuterium and a surface barrier or scintillation detector for the tritium ions.

  10. Technology benefits associated with accelerator production of tritium

    SciTech Connect (OSTI)

    Tuyle, G.J. van

    1998-12-31

    The Accelerator Production of Tritium (APT) offers a clean, safe, and reliable means of producing the tritium needed to maintain the nuclear deterrent. Tritium decays away naturally at a rate of about 5.5% per year; therefore, the tritium reservoirs in nuclear weapons must be periodically replenished. In recent years this has been accomplished by recycling tritium from weapons being retired from the stockpile. Although this strategy has served well since the last US tritium production reactor was shut down in 1988, a new tritium production capability will be required within ten years. Important technology benefits will result from direct utilization of some of the APT proton beam; others could result from advances in the technologies of particle accelerators and high power spallation targets. This report addresses those technology benefits.

  11. DEVELOPMENT OF THE BULK TRITIUM SHIPPING PACKAGING

    SciTech Connect (OSTI)

    Blanton, P.; Eberl, K.

    2008-09-14

    A new radioactive shipping packaging for transporting bulk quantities of tritium, the Bulk Tritium Shipping Package (BTSP), has been designed for the Department of Energy (DOE) as a replacement for a package designed in the early 1970s. This paper summarizes significant design features and describes how the design satisfies the regulatory safety requirements of the Code of Federal Regulations and the International Atomic Energy Agency. The BTSP design incorporates many improvements over its predecessor by implementing improved testing, handling, and maintenance capabilities, while improving manufacturability and incorporating new engineered materials. This paper also discusses the results from testing of the BTSP to 10 CFR 71 Normal Conditions of Transport and Hypothetical Accident Condition events. The programmatic need of the Department of Energy (DOE) to ship bulk quantities of tritium has been satisfied since the late 1970s by the UC-609 shipping package. The current Certificate of Conformance for the UC-609, USA/9932/B(U) (DOE), will expire in late 2011. Since the UC-609 was not designed to meet current regulatory requirements, it will not be recertified and thereby necessitates a replacement Type B shipping package for continued DOE tritium shipments in the future. A replacement tritium packaging called the Bulk Tritium Shipping Package (BTSP) is currently being designed and tested by Savannah River National Laboratory (SRNL). The BTSP consists of two primary assemblies, an outer Drum Assembly and an inner Containment Vessel Assembly (CV), both designed to mitigate damage and to protect the tritium contents from leaking during the regulatory Hypothetical Accident Condition (HAC) events and during Normal Conditions of Transport (NCT). During transport, the CV rests on a silicone pad within the Drum Liner and is covered with a thermal insulating disk within the insulated Drum Assembly. The BTSP packaging weighs approximately 500 lbs without contents and is 50-1/2 inches high by 24-1/2 inches in outside diameter. With contents the gross weight of the BTSP is 650 lbs. The BTSP is designed for the safe shipment of 150 grams of tritium in a solid or gaseous state. To comply with the federal regulations that govern Type B shipping packages, the BTSP is designed so that it will not lose tritium at a rate greater than the limits stated in 10CFR 71.51 of 10{sup -6} A2 per hour for the 'Normal Conditions of Transport' (NCT) and an A2 in 1 week under 'Hypothetical Accident Conditions' (HAC). Additionally, since the BTSP design incorporates a valve as part of the tritium containment boundary, secondary containment features are incorporated in the CV Lid to protect against gas leakage past the valve as required by 10CFR71.43(e). This secondary containment boundary is designed to provide the same level of containment as the primary containment boundary when subjected to the HAC and NCT criteria.

  12. Apparatus to recover tritium from tritiated molecules

    DOE Patents [OSTI]

    Swansiger, William A. (Livermore, CA)

    1988-01-01

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

  13. Recovery of tritium from tritiated molecules

    DOE Patents [OSTI]

    Swansiger, William A. (Livermore, CA)

    1987-01-01

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

  14. Overview of the Tritium research activities at Lawrence Livermore...

    Office of Environmental Management (EM)

    activities at LLNL More Documents & Publications Overview of Tritium Activities at the Laboratory for Laser Energetics NIF Presentation by Ed Moses EIS-0236-S1: Record of Decision...

  15. Tritium Formation and Mitigation in High-Temperature Reactor Systems

    SciTech Connect (OSTI)

    Piyush Sabharwall; Carl Stoots; Hans A. Schmutz

    2013-03-01

    Tritium is a radiologically active isotope of hydrogen. It is formed in nuclear reactors by neutron absorption and ternary fission events and can subsequently escape into the environment. To prevent the tritium contamination of proposed reactor buildings and surrounding sites, this study examines the root causes and potential mitigation strategies for permeation of tritium (such as: materials selection, inert gas sparging, etc...). A model is presented that can be used to predict permeation rates of hydrogen through metallic alloys at temperatures from 450–750 degrees C. Results of the diffusion model are presented for a steady production of tritium

  16. Tritium Formation and Mitigation in High-Temperature Reactors

    SciTech Connect (OSTI)

    Piyush Sabharwall; Carl Stoots

    2012-10-01

    Tritium is a radiologically active isotope of hydrogen. It is formed in nuclear reactors by neutron absorption and ternary fission events and can subsequently escape into the environment. To prevent the tritium contamination of proposed reactor buildings and surrounding sites, this study examines the root causes and potential mitigation strategies for permeation of tritium (such as: materials selection, inert gas sparging, etc...). A model is presented that can be used to predict permeation rates of hydrogen through metallic alloys at temperatures from 450–750 degrees C. Results of the diffusion model are presented for a steady production of tritium

  17. Test Results For Physical Separation Of Tritium From Noble Gases...

    Office of Environmental Management (EM)

    Test Results For Physical Separation Of Tritium From Noble Gases And It's Implications For Sensitivity And Accuracy In Air And Stack Monitoring Test Results For Physical Separation...

  18. Overview of Tritium Betavoltaic Power for Micro Sensors | Department...

    Office of Environmental Management (EM)

    Tritium Betavoltaic Power for Micro Sensors More Documents & Publications Studies on Lithium Manganese Rich MNC Composite Cathodes Vehicle Technologies Office Merit Review 2014:...

  19. Tritium Related Material Research -Irradiation Effect on Isotropic...

    Office of Environmental Management (EM)

    Related Material Research -Irradiation Effect on Isotropic Graphite Utilizing Heavy Ion-Irradiation- Tritium Related Material Research -Irradiation Effect on Isotropic Graphite...

  20. 2012 ACCOMPLISHMENTS - TRITIUM AGING STUDIES ON STAINLESS STEELS

    SciTech Connect (OSTI)

    Morgan, M.

    2013-01-31

    This report summarizes the research and development accomplishments during FY12 for the tritium effects on materials program. The tritium effects on materials program is designed to measure the long-term effects of tritium and its radioactive decay product, helium-3, on the structural properties of forged stainless steels which are used as the materials of construction for tritium reservoirs. The FY12 R&D accomplishments include: (1) Fabricated and Thermally-Charged 150 Forged Stainless Steel Samples with Tritium for Future Aging Studies; (2) Developed an Experimental Plan for Measuring Cracking Thresholds of Tritium-Charged-and-Aged Steels in High Pressure Hydrogen Gas; (3) Calculated Sample Tritium Contents For Laboratory Inventory Requirements and Environmental Release Estimates; (4) Published report on “Cracking Thresholds and Fracture Toughness Properties of Tritium-Charged-and-Aged Stainless Steels”; and, (5) Published report on “The Effects of Hydrogen, Tritium, and Heat Treatment on the Deformation and Fracture Toughness Properties of Stainless Steels”. These accomplishments are highlighted here and references given to additional reports for more detailed information.

  1. EFFECTS OF TRITIUM GAS EXPOSURE ON EPDM ELASTOMER

    SciTech Connect (OSTI)

    Clark, E.

    2009-12-11

    Samples of four formulations of ethylene-propylene diene monomer (EPDM) elastomer were exposed to initially pure tritium gas at one atmosphere and ambient temperature for various times up to about 420 days in closed containers. Two formulations were carbon-black-filled commercial formulations, and two were the equivalent formulations without filler synthesized for this work. Tritium effects on the samples were characterized by measuring the sample volume, mass, flexibility, and dynamic mechanical properties and by noting changes in appearance. The glass transition temperature was determined by analysis of the dynamic mechanical properties. The glass transition temperature increased significantly with tritium exposure, and the unfilled formulations ceased to behave as elastomers after the longest tritium exposure. The filled formulations were more resistant to tritium exposure. Tritium exposure made all samples significantly stiffer and therefore much less able to form a reliable seal when employed as O-rings. No consistent change of volume or density was observed; there was a systematic lowering of sample mass with tritium exposure. In addition, the significant radiolytic production of gas, mainly protium (H{sub 2}) and HT, by the samples when exposed to tritium was characterized by measuring total pressure in the container at the end of each exposure and by mass spectroscopy of a gas sample at the end of each exposure. The total pressure in the containers more than doubled after {approx}420 days tritium exposure.

  2. TRITIUM BARRIER MATERIALS AND SEPARATION SYSTEMS FOR THE NGNP

    SciTech Connect (OSTI)

    Sherman, S; Thad Adams, T

    2008-07-17

    Contamination of downstream hydrogen production plants or other users of high-temperature heat is a concern of the Next Generation Nuclear Plant (NGNP) Project. Due to the high operating temperatures of the NGNP (850-900 C outlet temperature), tritium produced in the nuclear reactor can permeate through heat exchangers to reach the hydrogen production plant, where it can become incorporated into process chemicals or the hydrogen product. The concentration limit for tritium in the hydrogen product has not been established, but it is expected that any future limit on tritium concentration will be no higher than the air and water effluent limits established by the NRC and the EPA. A literature survey of tritium permeation barriers, capture systems, and mitigation measures is presented and technologies are identified that may reduce the movement of tritium to the downstream plant. Among tritium permeation barriers, oxide layers produced in-situ may provide the most suitable barriers, though it may be possible to use aluminized surfaces also. For tritium capture systems, the use of getters is recommended, and high-temperature hydride forming materials such as Ti, Zr, and Y are suggested. Tritium may also be converted to HTO in order to capture it on molecular sieves or getter materials. Counter-flow of hydrogen may reduce the flux of tritium through heat exchangers. Recommendations for research and development work are provided.

  3. 34th Tritium Focus Group Meeting, Idaho National Laboratory,...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Hydrogen Isotopes Continuum-scale Modeling of Hydrogen and Helium Bubble Growth in Metals Fusion Nuclear Science and Technology Program - Status and Plans for Tritium Research...

  4. Evaluation of In-Situ Tritium Transport Parameters for Type 316...

    Office of Environmental Management (EM)

    In-Situ Tritium Transport Parameters for Type 316 Stainless Steel during Irradiation Evaluation of In-Situ Tritium Transport Parameters for Type 316 Stainless Steel during...

  5. Preliminary safety assessment for an IFE target fabrication facility

    SciTech Connect (OSTI)

    Latkowski, J F; Reyes, S; Besenbruch, G E; Goodin, D T

    2000-10-13

    We estimate possible ranges of tritium inventories for an inertial fusion energy (IFE) target fabrication facility producing various types of targets and using various production technologies. Target fill is the key subtask in determining the overall tritium inventory for the plant. By segmenting the inventory into multiple, parallel production lines--each with its own fill canister--and including an expansion tank to limit releases, we are able to ensure that a target fabrication facility would meet the accident dose goals of 10 mSv (1 rem) set forth in the Department of Energy's Fusion Safety Standards. For indirect-drive targets, we calculate release fractions for elements from lithium to bismuth and show that nearly all elements meet the dose goal. Our work suggests directions for future R&D that will help reduce total tritium inventories and increase the flexibility of target fabrication facilities.

  6. A solute transport model calibration procedure as applied to a tritium plume in the Savannah River Plant F-Area, South Carolina 

    E-Print Network [OSTI]

    Edwards, David Arthur

    1988-01-01

    on extracting values of hydrogeologic parameters which are physically meaningful. Results of modeling with fitted, non-physical parameters are usually of little value and often misleading. The calibration procedure which is associated with the multidimensional.... Monitoring Wells. Source Loading. Geology. Surface Hydrology and Physiography. Physical Hydrogeology Tritium Dispersion and Distribution. . 4 . . . . . . 8 . . . . . 1 1 . . 12 . . . 15 . . . 20 31 . . . 34 . . . . . 37 . . . . . . 37...

  7. Effectiveness Monitoring Report, MWMF Tritium Phytoremediation Interim Measures.

    SciTech Connect (OSTI)

    Hitchcock, Dan; Blake, John, I.

    2003-02-10

    This report describes and presents the results of monitoring activities during irrigation operations for the calendar year 2001 of the MWMF Interim Measures Tritium Phytoremediation Project. The purpose of this effectiveness monitoring report is to provide the information on instrument performance, analysis of CY2001 measurements, and critical relationships needed to manage irrigation operations, estimate efficiency and validate the water and tritium balance model.

  8. Fermilab | Tritium at Fermilab | Ferry Creek Results

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity ofkandz-cm11 Outreach Home Room NewsInformation Current HABFES OctoberEvan Racah861MayArtQuestions forFeature photo Tritium

  9. Fusion reactor high vacuum pumping: Charcoal cryosorber tritium exposure results

    SciTech Connect (OSTI)

    Sedgley, D.W.; Walthers, C.R.; Jenkins, E.M. (Grumman Aerospace Corp., Bethpage, NY (United States))

    1991-01-01

    Recent experiments, have shown the practically of using activated charcoal (coconut charcoal) at 4{degrees}K to pump helium and hydrogen isotopes for a fusion reactor. Both speed and capacity for deuterium/helium and tritium/helium-3 mixtures were shown to be satisfactory. The long term effects of tritium on the charcoal/cement system developed by Grumman and LLNL were not known and a program was undertaken to see what, if any, effect long term tritium exposure has on the cryosorber. Several charcoal on aluminum test samples were subjected to six months exposure of tritium at approximately 77{degrees}K. The tritium was scanned several times with a residual gas analyzer and the speed-capacity performance of the samples was measured before, approximately half way through and after the exposure. Modest effects were noted which would not seriously restrict charcoal's use as a cryosorber for fusion reactor high vacuum pumping applications. 4 refs., 8 figs.

  10. Derivation of dose conversion factors for tritium

    SciTech Connect (OSTI)

    Killough, G. G.

    1982-03-01

    For a given intake mode (ingestion, inhalation, absorption through the skin), a dose conversion factor (DCF) is the committed dose equivalent to a specified organ of an individual per unit intake of a radionuclide. One also may consider the effective dose commitment per unit intake, which is a weighted average of organ-specific DCFs, with weights proportional to risks associated with stochastic radiation-induced fatal health effects, as defined by Publication 26 of the International Commission on Radiological Protection (ICRP). This report derives and tabulates organ-specific dose conversion factors and the effective dose commitment per unit intake of tritium. These factors are based on a steady-state model of hydrogen in the tissues of ICRP's Reference Man (ICRP Publication 23) and equilibrium of specific activities between body water and other tissues. The results differ by 27 to 33% from the estimate on which ICRP Publication 30 recommendations are based. The report also examines a dynamic model of tritium retention in body water, mineral bone, and two compartments representing organically-bound hydrogen. This model is compared with data from human subjects who were observed for extended periods. The manner of combining the dose conversion factors with measured or model-predicted levels of contamination in man's exposure media (air, drinking water, soil moisture) to estimate dose rate to an individual is briefly discussed.

  11. TRITIUM TRANSPORT IN POLOIDAL FLOWS OF A DCLL BLANKET M.J. Pattison1

    E-Print Network [OSTI]

    Abdou, Mohamed

    ) blanket, tritium losses from the PbLi into cooling helium streams may occur when the liquid-metal breeder tritium from helium. The second is that tritium can make its way from the helium stream breeder. A proper understanding of the behavior of the PbLi flows and the production of tritium is vital

  12. Standard test method for nondestructive assay of plutonium, tritium and 241 Am by calorimetric assay

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2009-01-01

    Standard test method for nondestructive assay of plutonium, tritium and 241 Am by calorimetric assay

  13. Tritium and neutron measurements from deuterated Pd-Si

    SciTech Connect (OSTI)

    Claytor, T.N.; Tuggle, D.G.; Menlove, H.O.; Seeger, P.A.; Doty, W.R.; Rohwer, R.K. (Los Alamos National Laboratory, Los Alamos, New Mexico 87545 (United States))

    1991-05-10

    Evidence has been found for tritium and neutron production in palladium and silicon stacks when pulsed with a high electric current. These palladium-silicon stacks consist of alternating layers of pressed palladium and silicon powder. A pulsed high electric current is thought to promote non-equilibrium conditions important for tritium and neutron production. More than 2000 hours of neutron counting time has been accumulated in a underground, low background, environment with high efficiency counters (21%). Neutron emission has occurred as infrequent bursts or as low level emission lasting for up to 20 hours. In eight of 30 cells, excess tritium greater than 3 sigma has been observed. In each of these measurements, with the powder system, the ratio of tritium detected to total integrated total neutrons inferred has been anomalously high. Recent cells have shown reproducible tritium generation at a level of about 0.5 nCi/hr. Several hydrogen and air control cells have been run with no anomalous excess tritium or neutron emission above background. A singificant amount of the total palladium inventory (18%) has been checked for tritium contamination by three independent means.

  14. Tritium and neutron measurements from deuterated Pd-Si

    SciTech Connect (OSTI)

    Claytor, T.N.; Tuggle, D.G.; Menlove, H.O.; Seeger, P.A.; Doty, W.R.; Rohwer, R.K.

    1990-01-01

    Evidence has been found for tritium and neutron production in palladium and silicon stacks when pulsed with a high electric current. These palladium-silicon stacks consist of alternating layers of pressed palladium and silicon powder. A pulsed high electric current is thought to promote non equilibrium conditions important for tritium and neutron production. More than 2000 hours of neutron counting time has been accumulated in a underground, low background, environment with high efficiency counters (21%). Neutron emission has occurred as infrequent burst or as low level emission lasting for up to 20 hours. In eight of 30 cells, excess tritium greater than 3 sigma has been observed. In each of these measurements, with the powder system, the ratio of tritium detected to total integrated total neutrons inferred has been anomalously high. Recent cells have shown reproducible tritium generation at a level of about 0.5 nCi/hr. Several hydrogen and air control cells have been run with no anomalous excess tritium or neutron emission above background. A significant amount of the total palladium inventory (18%) has been checked for tritium contamination by three independent means. 12 refs., 6 figs., 2 tabs.

  15. Technology benefits resulting from accelerator production of tritium

    SciTech Connect (OSTI)

    NONE

    1998-12-31

    One of the early and most dramatic uses of nuclear transformations was in development of the nuclear weapons that brought World War II to an end. Despite that difficult introduction, nuclear weapons technology has been used largely as a deterrent to war throughout the latter half of the twentieth century. The Accelerator Production of Tritium (APT) offers a clean, safe, and reliable means of producing the tritium (a heavy form of hydrogen) needed to maintain the nuclear deterrent. Tritium decays away naturally at a rate of about 5.5% per year; therefore, the tritium reservoirs in nuclear weapons must be periodically replenished. In recent years this has been accomplished by recycling tritium from weapons being retired from the stockpile. Although this strategy has served well since the last US tritium production reactor was shut down in 1988, a new tritium production capability will be required within ten years. Some benefits will result from direct utilization of some of the APT proton beam; others could result from advances in the technologies of particle accelerators and high power spallation targets. The APT may save thousands of lives through the production of medical isotopes, and it may contribute to solving the nation`s problem in disposing of long-lived nuclear wastes. But the most significant benefit may come from advancing the technology, so that the great potential of accelerator applications can be realized during our lifetimes.

  16. MEASUREMENT OF TRITIUM DURING VOLOXIDATION OF ZIRCALOY-2 FUEL HULLS

    SciTech Connect (OSTI)

    Crowder, M.; Laurinat, J.; Stillman, J.

    2010-10-14

    A straightforward method to evaluate the tritium content of Zircaloy-2 cladding hulls via oxidation of the hull and capture of the volatilized tritium in liquids has been demonstrated. Hull samples were heated in air inside a thermogravimetric analyzer (TGA). The TGA was rapidly heated to 1000 C to oxidize the hulls and release absorbed tritium. To capture tritium, the TGA off-gas was bubbled through a series of liquid traps. The concentrations of tritium in bubbler solutions indicated that tritiated water vapor was captured nearly quantitatively. The average tritium content measured in the hulls was 19% of the amount of tritium produced by the fuel, according to ORIGEN2 isotope generation and depletion calculations. Published experimental data show that Zircaloy-2 oxidation follows an Arrhenius model, and that an initial, nonlinear oxidation rate is followed by a faster, linear rate after 'breakaway' of the oxide film. This study demonstrates that the linear oxidation rate of Zircaloy samples at 974 C is faster than predicted by the extrapolation of data from lower temperatures.

  17. Scientific and engineering services for the LANCE/ER accelerator production of tritium (APT) project

    SciTech Connect (OSTI)

    1994-12-05

    The APT project office is conducting a preconceptual design study for an accelerator driven concept to produce tritium. The facility will require new technology in many areas, since the scale of this accelerator is significantly larger then any in operation to date. The facility is composed of four subsystems: accelerator, target & blanket, balance of plant, and tritium purification system (TPS). New physics realms will be entered in order for the concept to be feasible; for example, extremely high energy levels of the entering protons that induce (multiplicative) spallation of the neutrons from the high Z target will occur. These are complex and require advance codes (MCNP) to predict the physics interactions and as well as deleterious material effects in the surrounding structures. Other issues include component cooling and complex thermal-hydraulics effects within the blanket and the beam {open_quotes}window.{close_quotes} In order to support a DOE mandated fast ROD schedule, Los Alamos APT staff will be provided with senior, engineering technical support staff with direct APT technology experience and whom are {open_quotes}on site{close_quotes}. This report contains resumes of the staff.

  18. Utilization of Kinetic Isotope Effects for the Concentration of Tritium

    SciTech Connect (OSTI)

    Brown, Gilbert M.; Meyer, Thomas j.; Moyer, Bruce A.

    1999-06-01

    The objective of this research program is to develop methods for concentrating tritium in water based on large primary isotope effects in catalytic redox processes. Basic research is being conducted to develop the chemistry of a complete cyclic process. Because tritium (generally present as HTO) is in a rapidly established equilibrium with protio-water, it moves with groundwater and separation from water cannot be achieved by the usual pump-and-treat methods using sorbants. The general methodology developed in this work will be applicable to a number of DOE waste streams, and as a consequence of the process tritium will be incorporated into an organic compound that will not readily exchange the tritium with groundwater. The process to be developed will remove tritium from H2O by concentrating it with respect to protio-water. This research involves developing chemical cycles that produce high concentration factors for HTO and T2O based on the discrimination of C-H and C-T bonds in oxidation reactions. Several steps are required in a cyclic process for the concentration of tritium in water. In the first step the tritium is incorporated in an organic compound. H-T discrimination occurs as the tritium containing compound is oxidized in a step involving a Ru(IV) oxo complex. Strong primary kinetic isotope effects lead to the oxidation of C-H bonds in preference to C-T bonds, and this reaction leads to concentration of tritium in the organic compound. The reduced form of the ruthenium compound can be reoxidized so that the oxidation step can be made catalytic.

  19. Apparatus for monitoring tritium in tritium contaminating environments using a modified Kanne chamber

    DOE Patents [OSTI]

    Anderson, David F. (Los Alamos, NM)

    1984-01-01

    A conventional Kanne tritium monitor has been redesigned to reduce its sensitivity to such contaminants as tritiated water vapor and tritiated oil. The high voltage electrode has been replaced by a wire cylinder and the collector electrode has been reduced in diameter. The area sensitive to contamination has thereby been reduced by about a factor of forty while the overall apparatus sensitivity and operation has not been affected. The design allows for in situ decontamination of the chambers, if necessary.

  20. Apparatus for monitoring tritium in tritium-contaminating environments using a modified Kanne chamber

    DOE Patents [OSTI]

    Anderson, D.F.

    1981-01-27

    A conventional Kanne tritium monitor has been redesigned to reduce its sensitivity to such contaminants as tritiated water vapor and tritiated oil. The high voltage electrode has been replaced by a wire cylinder and the collector electrode has been reduced in diameter. The area sensitive to contamination has thereby been reduced by about a factor of forty while the overall apparatus sensitivity and operation has not been affected. The design allows for in situ decontamination of the chambers, if necessary.

  1. Tritium Formation and Mitigation in High Temperature Reactors

    SciTech Connect (OSTI)

    Piyush Sabharwall; Carl Stoots

    2012-08-01

    Tritium is a radiologically active isotope of hydrogen. It is formed in nuclear reactors by neutron absorption and ternary fission events and can subsequently escape into the environment. In order to prevent the tritium contamination of proposed reactor buildings and surrounding sites, this paper examines the root causes and potential solutions for the production of this radionuclide, including materials selection and inert gas sparging. A model is presented that can be used to predict permeation rates of hydrogen through metallic alloys at temperatures from 450–750°C. Results of the diffusion model are presented for one steadystate value of tritium production in the reactor.

  2. Recommendations for Tritium Science and Technology Research and Development in Support of the Tritium Readiness Campaign, TTP-7-084

    SciTech Connect (OSTI)

    Senor, David J.

    2013-10-30

    Between 2006 and 2012 the Tritium Readiness Campaign Development and Testing Program produced significant advances in the understanding of in-reactor TPBAR performance. Incorporating these data into existing TPBAR performance models has improved permeation predictions, and the discrepancy between predicted and observed tritium permeation in the WBN1 coolant has been decreased by about 30%. However, important differences between predicted and observed permeation still remain, and there are significant knowledge gaps that hinder the ability to reliably predict other aspects of TPBAR performance such as tritium distribution, component integrity, and performance margins. Based on recommendations from recent Tritium Readiness Campaign workshops and reviews coupled with technical and programmatic priorities, high-priority activities were identified to address knowledge gaps in the near- (3-5 year), middle- (5-10 year), and long-term (10+ year) time horizons. It is important to note that there are many aspects to a well-integrated research and development program. The intent is not to focus exclusively on one aspect or another, but to approach the program in a holistic fashion. Thus, in addition to small-scale tritium science studies, ex-reactor tritium technology experiments such as TMED, and large-scale in-reactor tritium technology experiments such as TMIST, a well-rounded research and development program must also include continued analysis of WBN1 performance data and post-irradiation examination of TPBARs and lead use assemblies to evaluate model improvements and compare separate-effects and integral component behavior.

  3. On-line tritium production monitor

    DOE Patents [OSTI]

    Mihalczo, John T. (Oak Ridge, TN)

    1993-01-01

    A scintillation optical fiber system for the on-line monitoring of nuclear reactions in an event-by-event manner is described. In the measurement of tritium production one or more optical fibers are coated with enriched .sup.6 Li and connected to standard scintillation counter circuitry. A neutron generated .sup.6 Li(n )T reaction occurs in the coated surface of .sup.6 Li-coated fiber to produce energetic alpha and triton particles one of which enters the optical fiber and scintillates light through the fiber to the counting circuit. The coated optical fibers can be provided with position sensitivity by placing a mirror at the free end of the fibers or by using pulse counting circuits at both ends of the fibers.

  4. On-line tritium production monitor

    DOE Patents [OSTI]

    Mihalczo, J.T.

    1993-11-23

    A scintillation optical fiber system for the on-line monitoring of nuclear reactions in an event-by-event manner is described. In the measurement of tritium production one or more optical fibers are coated with enriched {sup 6}Li and connected to standard scintillation counter circuitry. A neutron generated {sup 6}Li(n)T reaction occurs in the coated surface of {sup 6}Li-coated fiber to produce energetic alpha and triton particles one of which enters the optical fiber and scintillates light through the fiber to the counting circuit. The coated optical fibers can be provided with position sensitivity by placing a mirror at the free end of the fibers or by using pulse counting circuits at both ends of the fibers. 5 figures.

  5. Ignition of deuterium-tritium fuel targets

    DOE Patents [OSTI]

    Musinski, D.L.; Mruzek, M.T.

    1991-08-27

    Disclosed is a method of igniting a deuterium-tritium ICF fuel target to obtain fuel burn in which the fuel target initially includes a hollow spherical shell having a frozen layer of DT material at substantially uniform thickness and cryogenic temperature around the interior surface of the shell. The target is permitted to free-fall through a target chamber having walls heated by successive target ignitions, so that the target is uniformly heated during free-fall to at least partially melt the frozen fuel layer and form a liquid single-phase layer or a mixed liquid/solid bi-phase layer of substantially uniform thickness around the interior shell surface. The falling target is then illuminated from exteriorly of the chamber while the fuel layer is at substantially uniformly single or bi-phase so as to ignite the fuel layer and release energy therefrom. 5 figures.

  6. Savannah River Tritium Enterprise exceeds productivity savings...

    National Nuclear Security Administration (NNSA)

    of a calibration process that will maintain the Measuring and Test Equipment tools for facility operations, while reducing the overall cost of the program (validated...

  7. Facility Environmental Vulnerability Assessment

    SciTech Connect (OSTI)

    Van Hoesen, S.D.

    2001-07-09

    From mid-April through the end of June 2001, a Facility Environmental Vulnerability Assessment (FEVA) was performed at Oak Ridge National Laboratory (ORNL). The primary goal of this FEVA was to establish an environmental vulnerability baseline at ORNL that could be used to support the Laboratory planning process and place environmental vulnerabilities in perspective. The information developed during the FEVA was intended to provide the basis for management to initiate immediate, near-term, and long-term actions to respond to the identified vulnerabilities. It was expected that further evaluation of the vulnerabilities identified during the FEVA could be carried out to support a more quantitative characterization of the sources, evaluation of contaminant pathways, and definition of risks. The FEVA was modeled after the Battelle-supported response to the problems identified at the High Flux Beam Reactor at Brookhaven National Laboratory. This FEVA report satisfies Corrective Action 3A1 contained in the Corrective Action Plan in Response to Independent Review of the High Flux Isotope Reactor Tritium Leak at the Oak Ridge National Laboratory, submitted to the Department of Energy (DOE) ORNL Site Office Manager on April 16, 2001. This assessment successfully achieved its primary goal as defined by Laboratory management. The assessment team was able to develop information about sources and pathway analyses although the following factors impacted the team's ability to provide additional quantitative information: the complexity and scope of the facilities, infrastructure, and programs; the significantly degraded physical condition of the facilities and infrastructure; the large number of known environmental vulnerabilities; the scope of legacy contamination issues [not currently addressed in the Environmental Management (EM) Program]; the lack of facility process and environmental pathway analysis performed by the accountable line management or facility owner; and poor facility and infrastructure drawings. The assessment team believes that the information, experience, and insight gained through FEVA will help in the planning and prioritization of ongoing efforts to resolve environmental vulnerabilities at UT-Battelle--managed ORNL facilities.

  8. Assessment of tritium in the Savannah River Site environment

    SciTech Connect (OSTI)

    Carlton, W.H.; Murphy, C.E. Jr.; Bauer, L.R. [and others

    1993-10-01

    This report is the first revision to a series of reports on radionuclides inn the SRS environment. Tritium was chosen as the first radionuclide in the series because the calculations used to assess the dose to the offsite population from SRS releases indicate that the dose due to tritium, through of small consequence, is one of the most important the radionuclides. This was recognized early in the site operation, and extensive measurements of tritium in the atmosphere, surface water, and ground water exist due to the effort of the Environmental Monitoring Section. In addition, research into the transport and fate of tritium in the environment has been supported at the SRS by both the local Department of Energy (DOE) Office and DOE`s Office of Health and Environmental Research.

  9. Utilization of Kinetic Isotope Effects for the Concentration of Tritium

    SciTech Connect (OSTI)

    Brown, Gilbert M.; Meyer, Thomas J.; Moyer, Bruce A.

    2000-06-01

    Work is in progress to develop methods for concentrating tritium in water based on large primary isotope effects in catalytic redox processes. Basic research is being conducted to develop the chemistry of a complete cyclic process. The process will remove tritium from H2O by concentrating it with respect to protio-water. This research involves developing chemical cycles that produce high concentration factors for HTO based on the discrimination of CH and C-T bonds in oxidation reactions. Several steps are required in a cyclic process for the concentration of tritium in water. In the first step, the tritium is incorporated in an organic compound. H-T discrimination occurs as the tritium containing compound is oxidized in a step involving a Ru(IV) oxo complex. Strong primary kinetic isotope effects lead to the oxidation of C-H bonds in preference to C-T bonds, and this reaction leads to concentration of tritium in the organic compound. The reduced form of the ruthenium compound can be reoxidized so that the oxidation step can be made catalytic.

  10. Low-energy beta spectroscopy using pin diodes to monitor tritium surface contamination

    SciTech Connect (OSTI)

    Wampler, W.R.; Doyle, B.L.

    1994-06-01

    We show that tritium betas emitted from a surface can be counted using a pin photodiode as a solid state charged particle detector. Furthermore, we show that the range of tritium betas through air is sufficient to allow measurement of tritium on samples in air by this method. These two findings make possible a new method to survey tritium surface contamination which has advantages over existing methods. We have built and tested several prototype instruments which use this method to measure tritium surface contamination, including a compact portable unit. The design of these instruments and results from tests and calibrations are described. Potential applications of this new method to monitor tritium are discussed.

  11. A PROTOTYPE FOUR INCH SHORT HYDRIDE (FISH) BED AS A REPLACEMENT TRITIUM STORAGE BED

    SciTech Connect (OSTI)

    Klein, J.; Estochen, E.; Shanahan, K.; Heung, L.

    2011-02-23

    The Savannah River Site (SRS) tritium facilities have used 1st generation (Gen1) metal hydride storage bed assemblies with process vessels (PVs) fabricated from 3 inch nominal pipe size (NPS) pipe to hold up to 12.6 kg of LaNi{sub 4.25}Al{sub 0.75} metal hydride for tritium gas absorption, storage, and desorption for over 15 years. The 2nd generation (Gen2) of the bed design used the same NPS for the PV, but the added internal components produced a bed nominally 1.2 m long, and presented a significant challenge for heater cartridge replacement in a footprint limited glove-box. A prototype 3rd generation (Gen3) metal hydride storage bed has been designed and fabricated as a replacement candidate for the Gen2 storage bed. The prototype Gen3 bed uses a PV pipe diameter of 4 inch NPS so the bed length can be reduced below 0.7 m to facilitate heater cartridge replacement. For the Gen3 prototype bed, modeling results show increased absorption rates when using hydrides with lower absorption pressures. To improve absorption performance compared to the Gen2 beds, a LaNi{sub 4.15}Al{sub 0.85} material was procured and processed to obtain the desired pressure-composition-temperature (PCT) properties. Other bed design improvements are also presented.

  12. RADIOLYTIC GAS PRODUCTION RATES OF POLYMERS EXPOSED TO TRITIUM GAS

    SciTech Connect (OSTI)

    Clark, E.

    2013-08-31

    Data from previous reports on studies of polymers exposed to tritium gas is further analyzed to estimate rates of radiolytic gas production. Also, graphs of gas release during tritium exposure from ultrahigh molecular weight polyethylene (UHMW-PE), polytetrafluoroethylene (PTFE, a trade name is Teflon®), and Vespel® polyimide are re-plotted as moles of gas as a function of time, which is consistent with a later study of tritium effects on various formulations of the elastomer ethylene-propylene-diene monomer (EPDM). These gas production rate estimates may be useful while considering using these polymers in tritium processing systems. These rates are valid at least for the longest exposure times for each material, two years for UHMW-PE, PTFE, and Vespel®, and fourteen months for filled and unfilled EPDM. Note that the production “rate” for Vespel® is a quantity of H{sub 2} produced during a single exposure to tritium, independent of length of time. The larger production rate per unit mass for unfilled EPDM results from the lack of filler- the carbon black in filled EPDM does not produce H{sub 2} or HT. This is one aspect of how inert fillers reduce the effects of ionizing radiation on polymers.

  13. Tritium production from a plasma discharge on palladium

    SciTech Connect (OSTI)

    Claytor, T.N.; Tuggle, D.G.; Jackson, D.D.

    1995-12-01

    Over the past year we have been able to demonstrate that a plasma loading method produces an exciting and unexpected amount of tritium. In contrast to electrochemical hydrogen or deuterium loading of palladium, this method yields a reproducible tritium generation rate when various electrical and physical conditions are met. We will show tritium generation rates for deuterium-palladium foreground runs that are up to 25 times larger than hydrogen-palladium control experiments using materials from the same batch. The reproducibility of the technique and the large signal to noise over background has allowed us to vary parameters that have been difficult to investigate with previous methods. We intend to illustrate the difference between batches of annealed palladium and as received palladium from several batches to demonstrate that the tritium generation rate can vary by a factor of 40 from batch to batch. The effect of other metals, wire and plate thicknesses on the tritium generation rate will be shown. We plan to discuss these new procedures, present typical results, and speculate concerning the implications for further work.

  14. Evaluation of In-Situ Tritium Transport Parameters for Type 316...

    Office of Environmental Management (EM)

    In-Situ Tritium Transport Parameters for Type 316 Stainless Steel during Irradiation D.J. Senor, W.G. Luscher K.K. Clayton, G.R. Longhurst Tritium Focus Group Meeting Savannah...

  15. EIS-0288: Production of Tritium in a Commercial Light Water Reactor

    Office of Energy Efficiency and Renewable Energy (EERE)

    This Environmental Impact Statement for the Production of Tritium in a Commercial Light Water Reactor (CLWR EIS) evaluates the environmental impacts associated with producing tritium at one or more...

  16. Continuous production of tritium in an isotope-production reactor with a separate circulation system

    DOE Patents [OSTI]

    Cawley, W.E.; Omberg, R.P.

    1982-08-19

    A method is described for producing tritium in a fast breeder reactor cooled with liquid metal. Lithium is allowed to flow through the reactor in separate loops in order to facilitate the production and removal of tritium.

  17. Validation of Hydrogen Exchange Methodology on Molecular Sieves for Tritium Removal from Contaminated Water

    Broader source: Energy.gov [DOE]

    Presentation from the 34th Tritium Focus Group Meeting held in Idaho Falls, Idaho on September 23-25, 2014.

  18. Methods for Post Irradiation Examination of Tritium Producing Burnable Absorber Rods

    Office of Energy Efficiency and Renewable Energy (EERE)

    Presentation from the 32nd Tritium Focus Group Meeting held in Germantown, Maryland on April 23-25, 2013.

  19. Improved Monitor Design and Configuration for Reducing Reported Tritium Discharges from the Orphee Research Reactor

    Broader source: Energy.gov [DOE]

    Presentation from the 33rd Tritium Focus Group Meeting held in Aiken, South Carolina on April 22-24, 2014.

  20. Thermal Removal of Tritium from Concrete and Soil to Reduce Groundwater Impacts

    Broader source: Energy.gov [DOE]

    Presentation from the 33rd Tritium Focus Group Meeting held in Aiken, South Carolina on April 22-24, 2014.

  1. Detection Limit of H and D for Tritium Process R&D

    Office of Energy Efficiency and Renewable Energy (EERE)

    Presentation from the 35th Tritium Focus Group Meeting held in Princeton, New Jersey on May 05-07, 2015.

  2. The Princeton Tritium Observatory for Light, Early Universe, Massive Neutrino Yield (PTOLEMY)

    Broader source: Energy.gov [DOE]

    Presentation from the 34th Tritium Focus Group Meeting held in Idaho Falls, Idaho on September 23-25, 2014.

  3. In-Reactor Measurement of Tritium Permeation through Stainless Steel Cladding

    Broader source: Energy.gov [DOE]

    Presentation from the 32nd Tritium Focus Group Meeting held in Germantown, Maryland on April 23-25, 2013.

  4. Evaluation of In-Situ Tritium Transport Parameters for Type 316 Stainless Steel during Irradiation

    Broader source: Energy.gov [DOE]

    Presentation from the 33rd Tritium Focus Group Meeting held in Aiken, South Carolina on April 22-24, 2014.

  5. Fusion Nuclear Science and Technology Program- Status and plans for tritium research

    Office of Energy Efficiency and Renewable Energy (EERE)

    Presentation from the 35th Tritium Focus Group Meeting held in Princeton, New Jersey on May 05-07, 2015.

  6. Fusion Nuclear Science and Technology Program- Status and Plans for Tritium Research

    Office of Energy Efficiency and Renewable Energy (EERE)

    Presentation from the 34th Tritium Focus Group Meeting held in Idaho Falls, Idaho on September 23-25, 2014.

  7. Tritium Related Material Research-Irradiation Effect on Isotropic Graphite Utilizing Heavy Ion-Irradiation-

    Broader source: Energy.gov [DOE]

    Presentation from the 34th Tritium Focus Group Meeting held in Idaho Falls, Idaho on September 23-25, 2014.

  8. Overview of the Tritium research activities at Lawrence Livermore National Laboratory (LLNL)

    Office of Energy Efficiency and Renewable Energy (EERE)

    Presentation from the 35th Tritium Focus Group Meeting held in Princeton, New Jersey on May 05-07, 2015.

  9. TRITIUM EFFECTS ON DYNAMIC MECHANICAL PROPERTIES OF POLYMERIC MATERIALS

    SciTech Connect (OSTI)

    Clark, E

    2008-11-12

    Dynamic mechanical analysis has been used to characterize the effects of tritium gas (initially 1 atm. pressure, ambient temperature) exposure over times up to 2.3 years on several thermoplastics-ultrahigh molecular weight polyethylene (UHMW-PE), polytetrafluoroethylene (PTFE), and Vespel{reg_sign} polyimide, and on several formulations of elastomers based on ethylene propylene diene monomer (EPDM). Tritium exposure stiffened the elastic modulus of UHMW-PE up to about 1 year and then softened it, and reduced the viscous response monotonically with time. PTFE initially stiffened, however the samples became too weak to handle after nine months exposure. The dynamic properties of Vespel{reg_sign} were not affected. The glass transition temperature of the EPDM formulations increased approximately 4 C. following three months tritium exposure.

  10. Tritium Permeability of Incoloy 800H and Inconel 617

    SciTech Connect (OSTI)

    Philip Winston; Pattrick Calderoni; Paul Humrickhouse

    2012-07-01

    Design of the Next Generation Nuclear Plant (NGNP) reactor and its high-temperature components requires information regarding the permeation of fission generated tritium and hydrogen product through candidate heat exchanger alloys. Release of fission-generated tritium to the environment and the potential contamination of the helium coolant by permeation of product hydrogen into the coolant system represent safety basis and product contamination issues. Of the three potential candidates for high-temperature components of the NGNP reactor design, only permeability for Incoloy 800H has been well documented. Hydrogen permeability data have been published for Inconel 617, but only in two literature reports and for partial pressures of hydrogen greater than one atmosphere, far higher than anticipated in the NGNP reactor. To support engineering design of the NGNP reactor components, the tritium permeability of Inconel 617 and Incoloy 800H was determined using a measurement system designed and fabricated at Idaho National Laboratory. The tritium permeability of Incoloy 800H and Inconel 617, was measured in the temperature range 650 to 950°C and at primary concentrations of 1.5 to 6 parts per million volume tritium in helium. (partial pressures of 10-6 atm)—three orders of magnitude lower partial pressures than used in the hydrogen permeation testing. The measured tritium permeability of Incoloy 800H and Inconel 617 deviated substantially from the values measured for hydrogen. This may be due to instrument offset, system absorption, presence of competing quantities of hydrogen, surface oxides, or other phenomena. Due to the challenge of determining the chemical composition of a mixture with such a low hydrogen isotope concentration, no categorical explanation of this offset has been developed.

  11. Tritium Permeability of Incoloy 800H and Inconel 617

    SciTech Connect (OSTI)

    Philip Winston; Pattrick Calderoni; Paul Humrickhouse

    2011-09-01

    Design of the Next Generation Nuclear Plant (NGNP) reactor and its high-temperature components requires information regarding the permeation of fission generated tritium and hydrogen product through candidate heat exchanger alloys. Release of fission-generated tritium to the environment and the potential contamination of the helium coolant by permeation of product hydrogen into the coolant system represent safety basis and product contamination issues. Of the three potential candidates for high-temperature components of the NGNP reactor design, only permeability for Incoloy 800H has been well documented. Hydrogen permeability data have been published for Inconel 617, but only in two literature reports and for partial pressures of hydrogen greater than one atmosphere, far higher than anticipated in the NGNP reactor. To support engineering design of the NGNP reactor components, the tritium permeability of Inconel 617 and Incoloy 800H was determined using a measurement system designed and fabricated at Idaho National Laboratory. The tritium permeability of Incoloy 800H and Inconel 617, was measured in the temperature range 650 to 950 C and at primary concentrations of 1.5 to 6 parts per million volume tritium in helium. (partial pressures of 10-6 atm) - three orders of magnitude lower partial pressures than used in the hydrogen permeation testing. The measured tritium permeability of Incoloy 800H and Inconel 617 deviated substantially from the values measured for hydrogen. This may be due to instrument offset, system absorption, presence of competing quantities of hydrogen, surface oxides, or other phenomena. Due to the challenge of determining the chemical composition of a mixture with such a low hydrogen isotope concentration, no categorical explanation of this offset has been developed.

  12. PPPL-2878 UC-426 February 1993 Tritium Diagnostics by Balmer-alpha Emission

    E-Print Network [OSTI]

    PPPL-2878 UC-426 February 1993 Tritium Diagnostics by Balmer-alpha Emission C H Skinner, A T Ramsey emission from tritium in a plasma may be distinguished from deuterium emission by a small isotope shift. A diagnostic system to measure tritium Balmer-alpha emission from the plasma edge has been installed on TFTR

  13. Uncertainty assessment and analysis of ITER in-VV tritium inventory determination

    SciTech Connect (OSTI)

    Cristescu, I. R.; Cristescu, I.; Glugla, M.; Murdoch, D.; Ciattaglia, S.

    2008-07-15

    Tracking of tritium inventories on ITER will be essential to ensure that the safety limits established for the mobilizable tritium inventory in the vacuum vessel are not violated. Tritium will be delivered to the ITER site from outside suppliers. Staring with the tritium imports the value of tritium inventory at ITER site will be known with a certain error that will propagate in time. During plasma operation, shot by shot measurements of the tritium delivered to the Torus and recovered will allow the amount of tritium trapped in the Torus to be computed at the end of the day. A case study for different measuring techniques and several measuring points for the tritium recovered from Torus have been done. An alternative method is to measure overnight the variation in the inventory of the storage and delivery system and the associated error when this method will be employed are presented. In order to reduce the errors on the tritium trapped in-vessel, at certain time intervals a method of global tritium inventory will be performed. The method envisages the transfer of all the mobilizable tritium from the plant and measurement of this inventory in the self-assay beds from the storage and delivery system. Evaluation of the most important sources of error for the tritium trapped in-vessel and means of minimization are eventually presented. (authors)

  14. Tritium handling experience in vacuum systems at TSTA (Tritium Systems Test Assembly)

    SciTech Connect (OSTI)

    Anderson, J.L.; Jenkins, E.M.; Walthers, C.R.; Yoshida, H.; Fukui, H.; Naruse, Y. (Los Alamos National Lab., NM (USA); Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan))

    1989-01-01

    Compound cryopumps have been added to the Tritium Systems Test Assembly (TSTA) integrated fusion fuel loop. Operations have been performed which closely simulate an actual fusion reactor pumping scenario. In addition, performance data have been taken that support the concept of using coconut charcoal as a sorbent at 4K for pumping helium. Later tests show that coconut charcoal may be used to co-pump D,T and He mixtures on a single 4K panel. Rotary spiral pumps have been used successfully in several applications at TSTA and have acquired more than 9000 hours of maintenance-free operation. Metal bellows pumps have been used to back the spiral pumps and have been relatively trouble free in loop operations. Bellows pumps also have more than 9000 hours of maintenance-free operation. 5 refs., 6 figs.

  15. Comparison of the recently proposed super-Marx generator approach to thermonuclear ignition with the deuterium-tritium laser fusion-fission hybrid concept by the Lawrence Livermore National Laboratory

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Winterberg, F.

    2009-10-29

    The recently proposed super-Marx generator pure deuterium microdetonation ignition concept is compared to the Lawrence Livermore National Ignition Facility (NIF) Laser deuterium-tritium fusion-fission hybrid concept (LIFE). In a super-Marx generator, a large number of ordinary Marx generators charge up a much larger second stage ultrahigh voltage Marx generator from which for the ignition of a pure deuterium microexplosion an intense GeV ion beam can be extracted. Typical examples of the LIFE concept are a fusion gain of 30 and a fission gain of 10, making up a total gain of 300, with about ten times more energy released into fission as compared to fusion. This means the substantial release of fission products, as in fissionless pure fission reactors. In the super-Marx approach for the ignition of pure deuterium microdetonation, a gain of the same magnitude can, in theory, be reached. If feasible, the super-Marx generator deuterium ignition approach would make lasers obsolete as a means for the ignition of thermonuclear microexplosions.

  16. Comparison of the recently proposed super-Marx generator approach to thermonuclear ignition with the deuterium-tritium laser fusion-fission hybrid concept by the Lawrence Livermore National Laboratory

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Winterberg, F.

    2009-01-01

    The recently proposed super-Marx generator pure deuterium microdetonation ignition concept is compared to the Lawrence Livermore National Ignition Facility (NIF) Laser deuterium-tritium fusion-fission hybrid concept (LIFE). In a super-Marx generator, a large number of ordinary Marx generators charge up a much larger second stage ultrahigh voltage Marx generator from which for the ignition of a pure deuterium microexplosion an intense GeV ion beam can be extracted. Typical examples of the LIFE concept are a fusion gain of 30 and a fission gain of 10, making up a total gain of 300, with about ten times more energy released into fissionmore »as compared to fusion. This means the substantial release of fission products, as in fissionless pure fission reactors. In the super-Marx approach for the ignition of pure deuterium microdetonation, a gain of the same magnitude can, in theory, be reached. If feasible, the super-Marx generator deuterium ignition approach would make lasers obsolete as a means for the ignition of thermonuclear microexplosions.« less

  17. Comparison of the recently proposed super-Marx generator approach to thermonuclear ignition with the deuterium-tritium laser fusion-fission hybrid concept by the Lawrence Livermore National Laboratory

    SciTech Connect (OSTI)

    Winterberg, F.

    2009-01-01

    The recently proposed super-Marx generator pure deuterium microdetonation ignition concept is compared to the Lawrence Livermore National Ignition Facility (NIF) Laser deuterium-tritium fusion-fission hybrid concept (LIFE). In a super-Marx generator, a large number of ordinary Marx generators charge up a much larger second stage ultrahigh voltage Marx generator from which for the ignition of a pure deuterium microexplosion an intense GeV ion beam can be extracted. Typical examples of the LIFE concept are a fusion gain of 30 and a fission gain of 10, making up a total gain of 300, with about ten times more energy released into fission as compared to fusion. This means the substantial release of fission products, as in fissionless pure fission reactors. In the super-Marx approach for the ignition of pure deuterium microdetonation, a gain of the same magnitude can, in theory, be reached. If feasible, the super-Marx generator deuterium ignition approach would make lasers obsolete as a means for the ignition of thermonuclear microexplosions.

  18. Solubility of hydrogen, deuterium, and tritium in palladium metal

    SciTech Connect (OSTI)

    Powell, G.L.; Laesser, R.

    1988-12-20

    The solubility of hydrogen, deuterium, and tritium in palladium has been measured, described analytically over a wide temperature range (296 to 1460 K), and used to calculate enthalpies of reaction, isotope separation factors, and phase boundaries. 44 refs., 11 figs., 5 tabs.

  19. Applications developed for byproduct /sup 85/Kr and tritium

    SciTech Connect (OSTI)

    Remini, W.C.; Case, F.N.; Haff, K.W.; Tiegs, S.M.

    1983-01-01

    The radionuclides, krypton-85 and tritium, both of which are gases under ordinary conditions, are used in many applications in industries and by the military forces. Krypton-85 is produced during the fissioning of uranium and is released during the dissolution of spent-fuel elements. It is a chemically inert gas that emits 0.695-MeV beta rays and a small yield of 0.54-MeV gammas over a half life of 10.3 years. Much of the /sup 85/Kr currently produced is released to the atmosphere; however, large-scale reprocessing of fuel will require collection of the gas and storage as a waste product. An alternative to storage is utilization, and since the chemical and radiation characteristics of /sup 85/Kr make this radionuclide a relatively low hazard from the standpoint of contamination and biological significance, a number of uses have been developed. Tritium is produced as a byproduct of the nuclear-weapons program, and it has a half life of 12.33 years. It has a 0.01861-MeV beta emission and no gamma emission. The absence of a gamma-ray energy eliminates the need for external shielding of the devices utilizing tritium, thus making them easily transportable. Many of the applications require only small quantities of /sup 85/Kr or tritium; however, these uses are important to the technology base of the nation. A significant development that has the potential for beneficial utilization of large quantities of /sup 85/Kr and of tritium involves their use in the production of low-level lighting devices. Since these lights are free from external fuel supplies, have a long half life (> 10 years), are maintenance-free, reliable, and easily deployed, both military and civilian airfield-lighting applications are being studied.

  20. ISIS Facility: Facility Design Challenges

    E-Print Network [OSTI]

    McDonald, Kirk

    ISIS Facility: Facility Design Challenges Matt Fletcher Head, Design Division ISIS Department, FNAL #12;ISIS -- neutrons Diamond -- X-rays #12;#12;· Lifetime · Reliable Operation · Flexibility

  1. Scoping Analyses on Tritium Permeation to VHTR Integarted Industrial Application Systems

    SciTech Connect (OSTI)

    Chang H. Oh; Eung S. Kim

    2011-03-01

    Tritium permeation is a very important current issue in the very high temperature reactor (VHTR) because tritium is easily permeated through high temperature metallic surfaces. Tritium permeations in the VHTR-integrated systems were investigated in this study using the tritium permeation analysis code (TPAC) that was developed by Idaho National Laboratory (INL). The INL TPAC is a numerical tool that is based on the mass balance equations of tritium containing species and hydrogen (i.e. HT, H2, HTO, HTSO4, TI) coupled with a variety of tritium sources, sink, and permeation models. In the TPAC, ternary fission and thermal neutron caption reactions with 6Li, 7Li 10B, 3He were taken into considerations as tritium sources. Purification and leakage models were implemented as main tritium sinks. Permeation of tritium and H2 through pipes, vessels, and heat exchangers were considered as main tritium transport paths. In addition, electroyzer and isotope exchange models were developed for analyzing hydrogen production systems including high temperature electrolysis (HTSE) and sulfur-iodine processes.

  2. DEVELOPMENT AND USE OF A BULK TRITIUM SHIPPING PACKAGE

    SciTech Connect (OSTI)

    Blanton, P.

    2010-09-30

    A shipping package for transporting tritium has been developed for use by the National Nuclear Safety Administration as a replacement for the DOE Model UC-609, a tritium package developed and used by the DOE and NRC since the early 1970s. This paper presents the major design features and highlights the improvements made over its predecessor by incorporating new engineered materials and implementing improved testing, handling, and maintenance capabilities, while improving manufacturability. A discussion will be provided demonstrating how the BTSP complies with the regulatory safety requirements of the Nuclear Regulatory Commission. The paper further summarizes the results of testing to 10 CFR 71 Normal Conditions of Transport and Hypothetical Accident Conditions events. Planned and possible future missions for this packaging will be addressed.

  3. FINITE ELEMENT ANALYSIS OF BULK TRITIUM SHIPPING PACKAGE

    SciTech Connect (OSTI)

    Jordan, J.

    2010-06-02

    The Bulk Tritium Shipping Package was designed by Savannah River National Laboratory. This package will be used to transport tritium. As part of the requirements for certification, the package must be shown to meet the scenarios of the Hypothetical Accident Conditions (HAC) defined in Code of Federal Regulations Title 10 Part 71 (10CFR71). The conditions include a sequential 30-foot drop event, 30-foot dynamic crush event, and a 40-inch puncture event. Finite Element analyses were performed to support and expand upon prototype testing. Cases similar to the tests were evaluated. Additional temperatures and orientations were also examined to determine their impact on the results. The peak stress on the package was shown to be acceptable. In addition, the strain on the outer drum as well as the inner containment boundary was shown to be acceptable. In conjunction with the prototype tests, the package was shown to meet its confinement requirements.

  4. Page 1 D. Dilling, T. Brown FIRE FACILITIES AND SITE REQUIREMENTS

    E-Print Network [OSTI]

    with regulations for nuclear facilities. They must also include systems to manage tritium and tritiated water, activated dust, and radioactive waste material. Maintenance activities on FIRE will require the use service connections will be required to feed power to the copper magnet system and deliver plasma

  5. IN-LINE CHEMICAL SENSOR DEPLOYMENT IN A TRITIUM PLANT

    SciTech Connect (OSTI)

    Tovo, L.; Wright, J.; Torres, R.; Peters, B.

    2013-10-02

    The Savannah River Tritium Plant (TP) relies on well understood but aging sensor technology for process gas analysis. Though new sensor technologies have been brought to various readiness levels, the TP has been reluctant to install technologies that have not been tested in tritium service. This gap between sensor technology development and incorporating new technologies into practical applications demonstrates fundamental challenges that exist when transitioning from status quo to state-of-the-art in an extreme environment such as a tritium plant. These challenges stem from three root obstacles: 1) The need for a comprehensive assessment of process sensing needs and requirements; 2) The lack of a pick-list of process-compatible sensor technologies; and 3) The need to test technologies in a tritium-contaminated process environment without risking production. At Savannah River, these issues are being addressed in a two phase project. In the first phase, TP sensing requirements were determined by a team of process experts. Meanwhile, Savannah River National Laboratory sensor experts identified candidate technologies and related them to the TP processing requirements. The resulting roadmap links the candidate technologies to actual plant needs. To provide accurate assessments of how a candidate sensor technology would perform in a contaminated process environment, an instrument demonstration station was established within a TP glove box. This station was fabricated to TP process requirements and designed to handle high activity samples. The combination of roadmap and demonstration station provides the following assets: ? Creates a partnership between the process engineers and researchers for sensor selection, maturation, and insertion, ? Selects the right sensors for process conditions ? Provides a means for safely inserting new sensor technology into the process without risking production, and ? Provides a means to evaluate off normal occurrences where and when they occur. This paper discusses the process to identify and demonstrate new sensor technologies for the Savannah River TP.

  6. A statistical analysis of personnel contaminations in 200 Area facilities

    SciTech Connect (OSTI)

    Wagner, M.A.; Stoddard, D.H.

    1983-05-18

    This study determined the frequency statistics of personnel contaminations in 200 Area facilities. These statistics are utilized in probability calculations for contamination risks, and are part of an effort to provide reliable information for use in safety studies. Data for this analysis were obtained from the 200 Area and the Tritium Area Fault Tree Data Banks and were analyzed with the aid of the STATPAC computer code.

  7. Tritium permeation experiments using reduced activation ferritic/martensitic steel tube and erbium oxide coating

    SciTech Connect (OSTI)

    Takumi Chikada; Masashi Shimada; Robert Pawelko; Takayuki Terai; Takeo Muroga

    2013-09-01

    Low concentration tritium permeation experiments have been performed on uncoated F82H and Er2O3-coated tubular samples in the framework of the Japan-US TITAN collaborative program. Tritium permeability of the uncoated sample with 1.2 ppm tritium showed one order of magnitude lower than that with 100% deuterium. The permeability of the sample with 40 ppm tritium was more than twice higher than that of 1.2 ppm, indicating a surface contribution at the lower tritium concentration. The Er2O3-coated sample showed two orders of magnitude lower permeability than the uncoated sample, and lower permeability than that of the coated plate sample with 100% deuterium. It was also indicated that the memory effect of ion chambers in the primary and secondary circuits was caused by absorption of tritiated water vapor that was generated by isotope exchange reactions between tritium and surface water on the coating.

  8. Tritium and neutron measurements of a solid state cell

    SciTech Connect (OSTI)

    Claytor, T.N.; Seeger, P.A.; Rohwer, R.K.; Tuggle, D.G.; Doty, W.R.

    1989-01-01

    A solid state cold fusion'' cell was constructed to test for non-equilibrium fusion in a solid. The stimulus for the design was the hypothesis that the electrochemical surface layer in the Pons- Fleischmann cell could be replaced with a metal-insulator- semiconductor (MIS) barrier. Cells were constructed of alternating layers of palladium and silicon powders pressed into a ceramic form and exposed to deuterium gas at 110 psia resulting in a D/Pd ratio of 0.7. Pulses of current were passed through the cells to populate non-equilibrium states at the MIS barriers. One cell showed neutron activity and was found to have a large amount of tritium, other cells have produced tritium at a low rate consistent with neutron emission below the threshold of observability. The branching ratio for n/p has been about 1 {times} 10{sup {minus}9} in all the experiments where a substantial amount of tritium has been found. 11 refs., 9 figs., 2 tabs.

  9. Improved modelling of helium and tritium production for spallation targets

    E-Print Network [OSTI]

    S. Leray; A. Boudard; J. Cugnon; J. C. David; A. Kelic-Heil; D. Mancusi; M. V. Ricciardi

    2009-12-11

    Reliable predictions of light charged particle production in spallation reactions are important to correctly assess gas production in spallation targets. In particular, the helium production yield is important for assessing damage in the window separating the accelerator vacuum from a spallation target, and tritium is a major contributor to the target radioactivity. Up to now, the models available in the MCNPX transport code, including the widely used default option Bertini-Dresner and the INCL4.2-ABLA combination of models, were not able to correctly predict light charged particle yields. The work done recently on both the intranuclear cascade model INCL4, in which cluster emission through a coalescence process has been introduced, and on the de-excitation model ABLA allows correcting these deficiencies. This paper shows that the coalescence emission plays an important role in the tritium and $^3He$ production and that the combination of the newly developed versions of the codes, INCL4.5-ABLA07, now lead to good predictions of both helium and tritium cross sections over a wide incident energy range. Comparisons with other available models are also presented.

  10. Report on the evaluation of the tritium producing burnable absorber rod lead test assembly. Revision 1

    SciTech Connect (OSTI)

    1997-03-01

    This report describes the design and fabrication requirements for a tritium-producing burnable absorber rod lead test assembly and evaluates the safety issues associated with tritium-producing burnable absorber rod irradiation on the operation of a commercial light water reactor. The report provides an evaluation of the tritium-producing burnable absorber rod design and concludes that irradiation can be performed within U.S. Nuclear Regulatory Commission regulations applicable to a commercial pressurized light water reactor.

  11. Facility Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1996-10-24

    Establishes facility safety requirements related to: nuclear safety design, criticality safety, fire protection and natural phenomena hazards mitigation.

  12. Facility Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1995-11-16

    Establishes facility safety requirements related to: nuclear safety design, criticality safety, fire protection and natural phenomena hazards mitigation.

  13. Concentration and removal of tritium and/or deuterium from water contaminated with tritium and/or deuterium

    DOE Patents [OSTI]

    Meyer, Thomas J. (Chapel Hill, NC); Narula, Poonam M. (Carrboro, NC)

    2001-01-01

    Concentration of tritium and/or deuterium that is a contaminant in H.sub.2 O, followed by separation of the concentrate from the H.sub.2 O. Employed are certain metal oxo complexes, preferably with a metal from Group VIII. For instance, [Ru.sup.IV (2,2',6',2"-terpyridine)(2,2'-bipyridine)(O)](ClO.sub.4).sub.2 is very suitable.

  14. The Tritium Under-flow Study at the Savannah River Site

    SciTech Connect (OSTI)

    Hiergesell, Robert A.

    2008-01-15

    An issue of concern at the Savannah River Site (SRS) over the past 20 years is whether tritiated groundwater originating at SRS might be the cause of low levels of tritium measured in certain domestic wells in Georgia. Tritium activity levels in several domestic wells have been observed to occur at levels comparable to what is measured in rainfall in areas surrounding SRS. Since 1988, there has been speculation that tritiated groundwater from SRS could flow under the river and find its way into Georgia wells. A considerable effort was directed at assessing the likelihood of trans-river flow, and 44 wells have been drilled by the USGS and the Georgia Department of Natural Resources. Also, as part of the data collection and analysis, the USGS developed a numerical model during 1997-98 to assess the possibility for such trans-river flow to occur. The model represented the regional groundwater flow system surrounding the Savannah River Site (SRS) in seven layers corresponding to the underlying hydrostratigraphic units, which was regarded as sufficiently detailed to evaluate whether groundwater originating at SRS could possibly flow beneath the Savannah River into Georgia. The model was calibrated against a large database of water-level measurements obtained from wells on both sides of the Savannah River and screened in each of the hydrostratigraphic units represented within the model. The model results verified that the groundwater movement in all hydrostratigraphic units proceeds laterally toward the Savannah River from both South Carolina and Georgia, and discharges into the river. Once the model was calibrated, a particle-track analysis was conducted to delineate areas of potential trans-river flow. Trans-river flow can occur in either an eastward or westward direction. The model indicated that all locations of trans-river flow are restricted to the Savannah River's flood plain, where groundwater passes immediately prior to discharging into the river. Whether the trans-river flow is eastward or westward depends primarily on the position of the Savannah River as it meanders back and forth within the flood plain and is limited to narrow sections of land adjacent to the river. With respect to the only location of westward trans-river flow that has a recharge area within the SRS, the new evaluations of hypothetical pumping scenarios indicated that only a very slight impact is incurred, even under the most extreme groundwater extraction scenario. The updated model did not result in a significant change in the location of the recharge areas at SRS and the only impact was measured in slight changes in the travel times associated with the travel path. The median groundwater travel times for particles released under each of the 4 groundwater extraction scenarios ranged from 366 to 507 years while. Under the most extreme scenario, that under which SRS groundwater extraction is discontinued, the shortest travel time was reduced from 90 to 79 years. It should be emphasized that the groundwater transit times do not include the time required for groundwater to migrate vertically downward across the uppermost aquifer (i.e. at the recharge area), thus the actual groundwater travel times could be up to several decades longer than what was calculated in the model. The exhaustive evaluations that have been conducted indicates that it is highly unlikely that tritiated groundwater originating at the SRS could migrate into Georgia and explain the low tritium activity levels that were originally observed in certain domestic water supply wells. Considering that those wells were located at some distance (several km) from the Savannah River, a far more likely explanation is that tritiated rainfall infiltrated the subsurface and recharged the shallow aquifer within which the well was finished.

  15. Thermal Removal of Tritium from Concrete and Soil to Reduce Groundwate...

    Office of Environmental Management (EM)

    & Thomas Kmetz - Savannah River Nuclear Solutions SP01 Savannah River Site & D-Area Heavy Water Processing Moderator Processing Subunit & Impacts to Groundwater Tritium (pCi...

  16. Neutron Emission Spectroscopy of Fuel Ion Rotation and Fusion Power Components Demonstrated in the Trace Tritium Experiments at JET

    E-Print Network [OSTI]

    Neutron Emission Spectroscopy of Fuel Ion Rotation and Fusion Power Components Demonstrated in the Trace Tritium Experiments at JET

  17. Thermal Design of a Metal Hydride Storage Bed, Permitting Tritium Accountancy to 0.1% Resolution and Repeatability

    E-Print Network [OSTI]

    Thermal Design of a Metal Hydride Storage Bed, Permitting Tritium Accountancy to 0.1% Resolution and Repeatability

  18. Physics of Aquatic Systems II, 6. Tritium Universitt HeidelbergInstitut fr Umweltphysik Physics of Aquatic Systems II

    E-Print Network [OSTI]

    Aeschbach-Hertig, Werner

    Umweltphysik 9 Tritium production in thermonuclear (hydrogen) bombs Total production: ~ 240 EBq = 6.5 GCi = 230

  19. Prospects for Relic Neutrino Detection at PTOLEMY: Princeton Tritium Observatory for Light, Early-Universe, Massive-Neutrino Yield

    Broader source: Energy.gov [DOE]

    Presentation from the 32nd Tritium Focus Group Meeting held in Germantown, Maryland on April 23-25, 2013.

  20. Fuel Ion Ratio Measurements in NBI Heated Deuterium Tritium Fusion Plasmas at JET using Neutron Emission Spectrometry

    E-Print Network [OSTI]

    Fuel Ion Ratio Measurements in NBI Heated Deuterium Tritium Fusion Plasmas at JET using Neutron Emission Spectrometry

  1. EFFECTS OF ONE WEEK TRITIUM EXPOSURE ON EPDM ELASTOMER

    SciTech Connect (OSTI)

    Clark, E

    2007-06-07

    This report documents test results for the exposure of four formulations of EPDM (ethylene-propylene diene monomer) elastomer to tritium gas at one atmosphere for approximately one week and characterization of material property changes and changes to the exposure gas during exposure. All EPDM samples were provided by Los Alamos National Laboratory (LANL). Material properties that were characterized include mass, sample dimensions, appearance, flexibility, and dynamic mechanical properties. The glass transition temperature was determined by analysis of the dynamic mechanical property data per ASTM standards. No change of glass transition temperature due to the short tritium gas exposure was observed. Filled and unfilled formulations of Dupont{reg_sign} Nordel{trademark} 1440 had a slightly higher glass transition temperature than filled and unfilled formulations of Uniroyal{reg_sign} Royalene{reg_sign} 580H; filled formulations had the same glass transition as unfilled. The exposed samples appeared the same as before exposure--there was no evidence of discoloration, and no residue on stainless steel spacers contacting the samples during exposure was observed. The exposed samples remained flexible--all formulations passed a break test without failing. The unique properties of polymers make them ideal for certain components in gas handling systems. Specifically, the resiliency of elastomers is ideal for sealing surfaces, for example in valves. EPDM, initially developed in the 1960s, is a hydrocarbon polymer used extensively for sealing applications. EPDM is used for its excellent combination of properties including high/low-temperature resistance, radiation resistance, aging resistance, and good mechanical properties. This report summarizes initial work to characterize effects of tritium gas exposure on samples of four types of EPDM elastomer: graphite filled and unfilled formulations of Nordel{trademark} 1440 and Royalene{reg_sign} 580H.

  2. Tritium Reduction and Control in the Vacuum Vessel During TFTR Outage and Decommissioning *

    E-Print Network [OSTI]

    Tritium Reduction and Control in the Vacuum Vessel During TFTR Outage and Decommissioning * W nearly three years of D­T operations, TFTR underwent an extended outage during which large port covers of the torus, a three tier system was developed for the outage in order to reduce and control the free tritium

  3. Tritium Reduction and Control in the Vacuum Vessel During TFTR Outage and Decommissioning*

    E-Print Network [OSTI]

    Tritium Reduction and Control in the Vacuum Vessel During TFTR Outage and Decommissioning* W years of D-T operations, TFTR underwent an extended outage during which large port covers were removed, a three tier system was developed for the outage in order to reduce and control the free tritium

  4. Tritium recovery from tritiated water with a two-stage palladium membrane reactor

    SciTech Connect (OSTI)

    Birdsell, S.A.; Willms, R.S.

    1997-04-01

    A process to recover tritium from tritiated water has been successfully demonstrated at TSTA. The 2-stage palladium membrane reactor (PMR) is capable of recovering tritium from water without generating additional waste. This device can be used to recover tritium from the substantial amount of tritiated water that is expected to be generated in the International Thermonuclear Experimental Reactor both from torus exhaust and auxiliary operations. A large quantity of tritiated waste water exists world wide because the predominant method of cleaning up tritiated streams is to oxidize tritium to tritiated water. The latter can be collected with high efficiency for subsequent disposal. The PMR is a combined catalytic reactor/permeator. Cold (non-tritium) water processing experiments were run in preparation for the tritiated water processing tests. Tritium was recovered from a container of molecular sieve loaded with 2,050 g (2,550 std. L) of water and 4.5 g of tritium. During this experiment, 27% (694 std. L) of the water was processed resulting in recovery of 1.2 g of tritium. The maximum water processing rate for the PMR system used was determined to be 0.5 slpm. This correlates well with the maximum processing rate determined from the smaller PMR system on the cold test bench and has resulted in valuable scale-up and design information.

  5. Lithium aluminate/zirconium material useful in the production of tritium

    DOE Patents [OSTI]

    Cawley, W.E.; Trapp, T.J.

    1984-10-09

    A composition is described useful in the production of tritium in a nuclear reactor. Lithium aluminate particles are dispersed in a matrix of zirconium. Tritium produced by the reactor of neutrons with the lithium are absorbed by the zirconium, thereby decreasing gas pressure within capsules carrying the material.

  6. Lithium aluminate/zirconium material useful in the production of tritium

    DOE Patents [OSTI]

    Cawley, William E. (Richland, WA); Trapp, Turner J. (Richland, WA)

    1984-10-09

    A composition is described useful in the production of tritium in a nuclear eactor. Lithium aluminate particles are dispersed in a matrix of zirconium. Tritium produced by the reactor of neutrons with the lithium are absorbed by the zirconium, thereby decreasing gas pressure within capsules carrying the material.

  7. In-Situ Imaging and Quantification of Tritium Surface Contamination via Coherent Fiber Bundle

    SciTech Connect (OSTI)

    Charles A. Gentile; John J. Parker; Stewart J. Zweben

    2001-11-12

    Princeton Plasma Physics Laboratory (PPPL) has developed a method of imaging tritium on in-situ surfaces for the purpose of real-time data collection. This method expands upon a previous tritium imaging concept, also developed at PPPL. Enhancements include an objective lens coupled to the entry aperture of a coherent fiber optic (CFO) bundle, and a relay lens connecting the exit aperture of the fiber bundle to an intensifier tube and a charge-coupled device (CCD) camera. The system has been specifically fabricated for use in determining tritium concentrations on first wall materials. One potential complication associated with the development of D-T [deuterium-tritium] fueled fusion reactors is the deposition of tritium (i.e., co-deposited layer) on the surface of the primary wall of the vacuum vessel. It would be advantageous to implement a process to accurately determine tritium distribution on these inner surfaces. This fiber optic imaging device provides a highly practical method for determining the location, concentration, and activity of surface tritium deposition. In addition, it can be employed for detection of tritium ''hot-spots'' and ''hide-out'' regions present on the surfaces being imaged.

  8. Method and apparatus for extracting tritium and preparing radioactive waste for disposal

    DOE Patents [OSTI]

    Heung, L.K.

    1994-03-29

    Apparatus is described for heating an object such as a nuclear target bundle to release and recover hydrogen and contain the disposable residue for disposal. The apparatus comprises an inverted furnace, a sleeve/crucible assembly for holding and enclosing the bundle, conveying equipment for placing the sleeve onto the crucible and loading the bundle into the sleeve/crucible, a lift for raising the enclosed bundle into the furnace, and hydrogen recovery equipment including a trap and strippers, all housed in a containment having negative internal pressure. The crucible/sleeve assembly has an internal volume that is sufficient to enclose and hold the bundle before heating; the crucible's internal volume is sufficient by itself to hold and enclose the bundle's volume after heating. The crucible can then be covered and disposed of; the sleeve, on the other hand, can be reused. 4 figures.

  9. Method and apparatus for extracting tritium and preparing radioactive waste for disposal

    DOE Patents [OSTI]

    Heung, Leung K. (Aiken, SC)

    1994-01-01

    Apparatus for heating an object such as a nuclear target bundle to release and recover hydrogen and contain the disposable residue for disposal. The apparatus comprises an inverted furnace, a sleeve/crucible assembly for holding and enclosing the bundle, conveying equipment for placing the sleeve onto the crucible and loading the bundle into the sleeve/crucible, a lift for raising the enclosed bundle into the furnace, and hydrogen recovery equipment including a trap and strippers, all housed in a containment having negative internal pressure. The crucible/sleeve assembly has an internal volume that is sufficient to enclose and hold the bundle before heating; the crucible's internal volume is sufficient by itself to hold and enclose the bundle's volume after heating. The crucible can then be covered and disposed of; the sleeve, on the other hand, can be reused.

  10. Shock timing measurements and analysis in deuterium-tritium-ice layered capsule implosions on NIF

    SciTech Connect (OSTI)

    Robey, H. F.; Celliers, P. M.; Moody, J. D.; Sater, J.; Parham, T.; Kozioziemski, B.; Dylla-Spears, R.; Ross, J. S.; LePape, S.; Ralph, J. E.; Dewald, E. L.; Berzak Hopkins, L.; Kroll, J. J.; Yoxall, B. E.; Hamza, A. V.; Landen, O. L.; Edwards, M. J. [Lawrence Livermore National Laboratory, Livermore, California 94551 (United States)] [Lawrence Livermore National Laboratory, Livermore, California 94551 (United States); Hohenberger, M.; Boehly, T. R. [Laboratory for Laser Energetics, Rochester, New York 14623 (United States)] [Laboratory for Laser Energetics, Rochester, New York 14623 (United States); Nikroo, A. [General Atomics, San Diego, California 92196 (United States)] [General Atomics, San Diego, California 92196 (United States)

    2014-02-15

    Recent advances in shock timing experiments and analysis techniques now enable shock measurements to be performed in cryogenic deuterium-tritium (DT) ice layered capsule implosions on the National Ignition Facility (NIF). Previous measurements of shock timing in inertial confinement fusion implosions [Boehly et al., Phys. Rev. Lett. 106, 195005 (2011); Robey et al., Phys. Rev. Lett. 108, 215004 (2012)] were performed in surrogate targets, where the solid DT ice shell and central DT gas were replaced with a continuous liquid deuterium (D2) fill. These previous experiments pose two surrogacy issues: a material surrogacy due to the difference of species (D2 vs. DT) and densities of the materials used and a geometric surrogacy due to presence of an additional interface (ice/gas) previously absent in the liquid-filled targets. This report presents experimental data and a new analysis method for validating the assumptions underlying this surrogate technique. Comparison of the data with simulation shows good agreement for the timing of the first three shocks, but reveals a considerable discrepancy in the timing of the 4th shock in DT ice layered implosions. Electron preheat is examined as a potential cause of the observed discrepancy in the 4th shock timing.

  11. Derived Intervention Levels for Tritium Based on Food and Drug Administration Methodology Using ICRP 56 Dose Coefficients

    SciTech Connect (OSTI)

    Blanchard, A.

    1999-06-09

    In 1998, the FDA released its recommendations for age-dependent derived intervention levels for several radionuclides involved in nuclear accidents. One radionuclide that is not included in that document is tritium. Therefore an analysis is presented here using dose coefficients from ICRP 56 to develop Derived Intervention Levels (DILs) for tritium in two forms: water (HTO) and organically bound tritium (OBT).

  12. ACUTRI a computer code for assessing doses to the general public due to acute tritium releases

    E-Print Network [OSTI]

    Yokoyama, S; Noguchi, H; Ryufuku, S; Sasaki, T

    2002-01-01

    Tritium, which is used as a fuel of a D-T burning fusion reactor, is the most important radionuclide for the safety assessment of a nuclear fusion experimental reactor such as ITER. Thus, a computer code, ACUTRI, which calculates the radiological impact of tritium released accidentally to the atmosphere, has been developed, aiming to be of use in a discussion of licensing of a fusion experimental reactor and an environmental safety evaluation method in Japan. ACUTRI calculates an individual tritium dose based on transfer models specific to tritium in the environment and ICRP dose models. In this calculation it is also possible to analyze statistically on meteorology in the same way as a conventional dose assessment method according to the meteorological guide of the Nuclear Safety Commission of Japan. A Gaussian plume model is used for calculating the atmospheric dispersion of tritium gas (HT) and/or tritiated water (HTO). The environmental pathway model in ACUTRI considers the following internal exposures: i...

  13. NaNi sub 3 Mn sub 2 alloy as a tritium storage material

    SciTech Connect (OSTI)

    Ide, T.; Okuno, K.; Konishi, S.; Sakai, F.; Fukui, H.; Enoeda, M.; Naruse, Y.; Anderson, J.L.; Bartlit, J.R. (Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan); Los Alamos National Lab., NM (USA))

    1989-01-01

    An all metal apparatus has been constructed and installed in the main cell of the Tritium System Assembly (TSTA) at Los Alamos National Laboratory, as a separate experiment, to handle about 2600 Ci of tritium for study of metal tritides of potential application for storing tritium in fusion fuel processing. The apparatus is similar to that used for protium/deuterium gas but some modifications were made to assure safe handling of tritium. The pressure-composition isotherms for the LaNi{sub 3}Mn{sub 2}-protium (H), deuterium (D) and tritium (T) system were measured to study isotopic effects in the temperature range of 60 {degree}C to 250 {degree}C, the pressure range below 120 kPa. 2 refs., 10 figs.

  14. Tritium stripping in a nitrogen glove box using palladium/zeolite and SAES St 198{trademark}

    SciTech Connect (OSTI)

    Klien, J.E.; Wermer, J.R.

    1995-01-01

    Glove box clean-up experiments were conducted in a nitrogen glove box using palladium deposited on zeolite (Pd/z) and a SAES St 198{trademark} getter as tritium stripping materials. Protium/deuterium samples spiked with tritium were released into a 620 liter glove box to simulate tritium releases in a 10,500 liter glove box. The Pd/z and the SAES St 198{trademark} stripper beds produced a reduction in tritium activity of approximately two to three orders of magnitude and glove box clean-up was limited by a persistent background tritium activity level. Attempts to significantly reduce the glove box activity to lower levels without purging were unsuccessful.

  15. Tritium stripping in a nitrogen glove box using palladium/zeolite and SAES St 198

    SciTech Connect (OSTI)

    Klein, J.E.; Wermer, J.R.

    1995-10-01

    Glove box clean-up experiments were conducted in a nitrogen glove box using palladium deposited on zeolite (Pd/z) and a SAES St 198 getter as tritium stripping materials. Protium/deuterium samples spiked with tritium were released into a 620 liter glove box to simulate tritium releases in a 10,500 liter glove box. The Pd/z and the SAES St 198 stripper beds produced a reduction in tritium activity of approximately two to three orders of magnitude and glove box clean-up was limited by a persistent background tritium activity level. Attempts to significantly reduce the glove box activity to lower levels without purging were unsuccessful. 3 refs., 6 figs., 1 tab.

  16. Evaluation of Technologies to Complement/Replace Mass Spectrometers in the Tritium Facilities

    SciTech Connect (OSTI)

    Tovo, L. L.; Lascola, R. J.; Spencer, W. A.; McWhorter, C. S.; Zeigler, K. E.

    2005-08-30

    The primary goal of this work is to determine the suitability of the Infraran sensor for use in the Palladium Membrane Reactor. This application presents a challenge for the sensor, since the process temperature exceeds its designed operating range. We have demonstrated that large baseline offsets, comparable to the sensor response to the analyte, are obtained if cool air is blown across the sensor. We have also shown that there is a strong environmental component to the noise. However, the current arrangement does not utilize a reference detector. The strong correlation between the CO and H{sub 2}O sensor responses to environmental changes indicate that a reference detector can greatly reduce the environmental sensitivity. In fact, incorporation of a reference detector is essential for the sensor to work in this application. We have also shown that the two sensor responses are adequately independent. Still, there are several small corrections which must to be made to the sensor response to accommodate chemical and physical effects. Interactions between the two analytes will alter the relationship between number density and pressure. Temperature and pressure broadening will alter the relationship between absorbance and number density. The individual effects are small--on the order of a few percent or less--but cumulatively significant. Still, corrections may be made if temperature and total pressure are independently measured and incorporated into a post-analysis routine. Such corrections are easily programmed and automated and do not represent a significant burden for installation. The measurements and simulations described above indicate that with appropriate corrections, the Infraran sensor can approach the 1-1.5% measurement accuracy required for effective PMR process control. It is also worth noting that the Infraran may be suitable for other gas sensing applications, especially those that do not need to be made in a high-temperature environment. Any gas with an infrared absorption (methane, ammonia, etc.) may be detected so long as an appropriate bandpass filter can be manufactured. Note that homonuclear diatomic molecules (hydrogen and its isotopes, nitrogen, oxygen) do not have infrared absorptions. We have shown that the sensor response may be adequately predicted using commercially available software. Measurement of trace concentrations is limited by the broad spectral bandpass, since the total signal includes non-absorbed frequencies. However, cells with longer pathlengths can be designed to address this problem.

  17. Radiological Training for Tritium Facilities DOE-HDBK-1105-2002

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious RankADVANCED MANUFACTURINGEnergy BillsNo. 195 - Oct. 7,DOE HDBK-1113-2008 April 2008 DOE

  18. NNSA Breaks Ground on Tritium Facilities at SRS | National Nuclear Security

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity ofkandz-cm11 Outreach Home Room NewsInformationJessework usesof Energy Moving Basic NERSCKeyNuclear

  19. Facility Representatives

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2006-04-06

    REPLACED BY DOE-STD-1063 | SUPERSEDING DOE-STD-1063-2000 (MARCH 2000) The purpose of the DOE Facility Representative Program is to ensure that competent DOE staff personnel are assigned to oversee the day-to-day contractor operations at DOE’s hazardous nuclear and non-nuclear facilities.

  20. Facility Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2005-12-22

    This Order establishes facility and programmatic safety requirements for Department of Energy facilities, which includes nuclear and explosives safety design criteria, fire protection, criticality safety, natural phenomena hazards mitigation, and the System Engineer Program. Cancels DOE O 420.1A. DOE O 420.1B Chg 1 issued 4-19-10.

  1. Tritium Specific Adsorption Simulation Utilizing the OSPREY Model

    SciTech Connect (OSTI)

    Veronica Rutledge; Lawrence Tavlarides; Ronghong Lin; Austin Ladshaw

    2013-09-01

    During the processing of used nuclear fuel, volatile radionuclides will be discharged to the atmosphere if no recovery processes are in place to limit their release. The volatile radionuclides of concern are 3H, 14C, 85Kr, and 129I. Methods are being developed, via adsorption and absorption unit operations, to capture these radionuclides. It is necessary to model these unit operations to aid in the evaluation of technologies and in the future development of an advanced used nuclear fuel processing plant. A collaboration between Fuel Cycle Research and Development Offgas Sigma Team member INL and a NEUP grant including ORNL, Syracuse University, and Georgia Institute of Technology has been formed to develop off gas models and support off gas research. This report is discusses the development of a tritium specific adsorption model. Using the OSPREY model and integrating it with a fundamental level isotherm model developed under and experimental data provided by the NEUP grant, the tritium specific adsorption model was developed.

  2. SHINE Tritium Nozzle Design: Activity 6, Task 1 Report

    SciTech Connect (OSTI)

    Okhuysen, Brett S.; Pulliam, Elias Noel

    2015-11-05

    In FY14, we studied the qualitative and quantitative behavior of a SHINE/PNL tritium nozzle under varying operating conditions. The result is an understanding of the nozzle’s performance in terms of important flow features that manifest themselves under different parametric profiles. In FY15, we will consider nozzle design with a focus on nozzle geometry and integration. From FY14 work, we will understand how the SHINE/PNL nozzle behaves under different operating scenarios. The first task for FY15 is to evaluate the FY14 model as a predictor of the actual flow. Considering different geometries is more time-intensive than parameter studies, therefore we recommend considering any relevant flow features that were not included in the FY14 model. In the absence of experimental data, it is particularly important to consider any sources of heat in the domain or boundary conditions that may affect the flow and incorporate these into the simulation if they are significant. Additionally, any geometric features of the beamline segment should be added to the model such as the orifice plate. The FY14 model works with hydrogen. An improvement that can be made for FY15 is to develop CFD properties for tritium and incorporate those properties into the new models.

  3. Hydrogen, Deuterium and Tritium in Palladium: An Elastic Constants Study

    SciTech Connect (OSTI)

    Bach, H.T.; Schwarz, R.B.; Tuggle, D.G.

    2005-07-15

    We have used resonant ultrasound spectroscopy to measure the three independent elastic constants of Pd-H, Pd-D, and Pd-T single crystal at 300K as a function of hydrogen, deuterium, and tritium concentration, respectively. The addition of interstitial H (D, or T) atoms, located at (0,1/2,0) in the fcc Pd lattice, affects all three elastic constants C', C{sub 44}, and B. In the mixed ({alpha}+{beta}) phase, and with increasing H isotope, the shear modulus C' shows an abnormal softening whereas C{sub 44} and B do not. This is explained in terms of Zener-type an elastic relaxations affecting the shape of the hydride phases in the coherent({alpha}+{beta}) two-phase mixture In the single {beta}-phase, C' shows a strong isotope dependence whereas C{sub 44} and B show none. This behavior is explained in terms of differences in the excitation of optical phonons. In Pd-T, {sup 3}He is produced by the radioactive decay of tritium. We have measured in situ the swelling and the change in the elastic constants in Pd-T as a function of aging time. Aging ({sup 3}He formation) affects all three elastic constants. These measurements are being used to understand the early stages of {sup 3}H-{sup 3}He cluster formation in aged Pd-T crystal.

  4. Hydrogen, deuterium and tritium in palladium: An eleastic constants study

    SciTech Connect (OSTI)

    Bach, H. T.; Schwarz, R. B.; Tuggle, D. G.

    2004-01-01

    We have used resonant ultrasound spectroscopy to measure the three independent elastic constants of Pd-H, Pd-D, and Pd-T single crystal at 300K as a junction of hydrogen, deuterium, and tritium concentration, respectively. The addition of interstitial H (D, or T) atoms, located at (0, 1/2, 0) in the fcc Pd lattice, affects all three elastic constants C, C{sub 44}, and B. In the mixed (a+{beta}) phase, and with increasing H isotope, the shear modulus C' shows an abnormal softening whereas C{sub 44} and B do not. This is explained in terms of Zener-type anelastic relaxations affecting the shape of the hydride phases in the coherent ({alpha}+{beta}) two-phase mixture In the single {beta}-phase, C' shows a strong isotope dependence whereas C{sub 44} and B show none. This behavior is explained in terms of differences in the excitation of optical phonons. In Pd-T, {sup 3}He is produced by the radioactive decay of tritium. We have measured in situ the swelling and the change in the elastic constants in Pd-T as a function of aging time. Aging ({sup 3}He formation) affects all three elastic constants. These measurements are being used to understand the early stages of {sup 3}H-{sup 3}He clusterformation in aged Pd-T crystal.

  5. Development and Verification of Tritium Analyses Code for a Very High Temperature Reactor

    SciTech Connect (OSTI)

    Chang H. Oh; Eung S. Kim

    2009-09-01

    A tritium permeation analyses code (TPAC) has been developed by Idaho National Laboratory for the purpose of analyzing tritium distributions in the VHTR systems including integrated hydrogen production systems. A MATLAB SIMULINK software package was used for development of the code. The TPAC is based on the mass balance equations of tritium-containing species and a various form of hydrogen (i.e., HT, H2, HTO, HTSO4, and TI) coupled with a variety of tritium source, sink, and permeation models. In the TPAC, ternary fission and neutron reactions with 6Li, 7Li 10B, 3He were taken into considerations as tritium sources. Purification and leakage models were implemented as main tritium sinks. Permeation of HT and H2 through pipes, vessels, and heat exchangers were importantly considered as main tritium transport paths. In addition, electroyzer and isotope exchange models were developed for analyzing hydrogen production systems including both high-temperature electrolysis and sulfur-iodine process. The TPAC has unlimited flexibility for the system configurations, and provides easy drag-and-drops for making models by adopting a graphical user interface. Verification of the code has been performed by comparisons with the analytical solutions and the experimental data based on the Peach Bottom reactor design. The preliminary results calculated with a former tritium analyses code, THYTAN which was developed in Japan and adopted by Japan Atomic Energy Agency were also compared with the TPAC solutions. This report contains descriptions of the basic tritium pathways, theory, simple user guide, verifications, sensitivity studies, sample cases, and code tutorials. Tritium behaviors in a very high temperature reactor/high temperature steam electrolysis system have been analyzed by the TPAC based on the reference indirect parallel configuration proposed by Oh et al. (2007). This analysis showed that only 0.4% of tritium released from the core is transferred to the product hydrogen. The amount of tritium in the product hydrogen was estimated to be approximately an order less than the gaseous effluent limit for tritium.

  6. 5.10 Tritium Geochemistry And Kd Values 5.10.1 Overview: Important Aqueous-and Solid-Phase Parameters

    E-Print Network [OSTI]

    of processes such as thermal neutron reactions with 6 Li. As an isotope of hydrogen, tritium in soil systems processes affect the mobility of tritium in soil/water systems. 5.10.5 Adsorption/Desorption Because tritium, precipitation, and sorption processes are not expected to affect the mobility of tritium in soil/water systems

  7. RECOMMENDED TRITIUM OXIDE DEPOSITION VELOCITY FOR USE IN SAVANNAH RIVER SITE SAFETY ANALYSES

    SciTech Connect (OSTI)

    Lee, P.; Murphy, C.; Viner, B.; Hunter, C.; Jannik, T.

    2012-04-03

    The Defense Nuclear Facilities Safety Board (DNFSB) has recently questioned the appropriate value for tritium deposition velocity used in the MELCOR Accident Consequence Code System Ver. 2 (Chanin and Young 1998) code when estimating bounding dose (95th percentile) for safety analysis (DNFSB 2011). The purpose of this paper is to provide appropriate, defensible values of the tritium deposition velocity for use in Savannah River Site (SRS) safety analyses. To accomplish this, consideration must be given to the re-emission of tritium after deposition. Approximately 85% of the surface area of the SRS is forested. The majority of the forests are pine plantations, 68%. The remaining forest area is 6% mixed pine and hardwood and 26% swamp hardwood. Most of the path from potential release points to the site boundary is through forested land. A search of published studies indicate daylight, tritiated water (HTO) vapor deposition velocities in forest vegetation can range from 0.07 to 2.8 cm/s. Analysis of the results of studies done on an SRS pine plantation and climatological data from the SRS meteorological network indicate that the average deposition velocity during daylight periods is around 0.42 cm/s. The minimum deposition velocity was determined to be about 0.1 cm/s, which is the recommended bounding value. Deposition velocity and residence time (half-life) of HTO in vegetation are related by the leaf area and leaf water volume in the forest. For the characteristics of the pine plantation at SRS the residence time corresponding to the average, daylight deposition velocity is 0.4 hours. The residence time corresponding to the night-time deposition velocity of 0.1 cm/s is around 2 hours. A simple dispersion model which accounts for deposition and re-emission of HTO vapor was used to evaluate the impact on exposure to the maximally exposed offsite individual (MOI) at the SRS boundary (Viner 2012). Under conditions that produce the bounding, 95th percentile MOI exposure, i.e., low wind speed, weak turbulence, night, low deposition velocity, the effect of deposition and re-emission on MOI exposure was found to be very small. The exposure over the two hour period following arrival of the plume was found to be decreased by less than 0.05 %. Furthermore the sensitivity to deposition velocity was low. Increasing deposition velocity to 0.5 cm/s reduced exposure to 0.3 %. After a 24 hour period, an MOI would have been exposed to all of the released material. Based on the low sensitivity of MOI exposure to the value of deposition velocity when re-emission is considered, it is appropriately conservative to use a 0.0 cm/s effective deposition velocity for safety analysis in the MACCS2 code.

  8. Fast Flux Test Facility (FFTF) Briefing Book 1 Summary

    SciTech Connect (OSTI)

    WJ Apley

    1997-12-01

    This report documents the results of evaluations preformed during 1997 to determine what, if an, future role the Fast Flux Test Facility (FFTF) might have in support of the Department of Energy’s tritium productions strategy. An evaluation was also conducted to assess the potential for the FFTF to produce medical isotopes. No safety, environmental, or technical issues associated with producing 1.5 kilograms of tritium per year in the FFTF have been identified that would change the previous evaluations by the Department of Energy, the JASON panel, or Putnam, Hayes & Bartlett. The FFTF can be refitted and restated by July 2002 for a total expenditure of $371 million, with an additional $64 million of startup expense necessary to incorporate the production of medical isotopes. Therapeutic and diagnostic applications of reactor-generated medical isotopes will increase dramatically over the next decade. Essential medical isotopes can be produced in the FFTF simultaneously with tritium production, and while a stand-alone medical isotope mission for the facility cannot be economically justified given current marker conditions, conservative estimates based on a report by Frost &Sullivan indicate that 60% of the annual operational costs (reactor and fuel supply) could be offset by revenues from medical isotope production within 10 yeas of restart. The recommendation of the report is for the Department of Energy to continue to maintain the FFTF in standby and proceed with preparation of appropriate Nations Environmental Policy Act documentation in full consultation with the public to consider the FFTF as an interim tritium production option (1.5 kilograms/year) with a secondary mission of producing medical isotopes.

  9. Milestone report - M4FT-14OR0302102b - Evaluation of Tritium Content and Release from Surry-2 Fuel Cladding

    SciTech Connect (OSTI)

    Robinson, Sharon M.; Chattin, Marc Rhea; Giaquinto, Joseph M.; Jubin, Robert Thomas

    2014-09-01

    To design and operate future reprocessing plants in a safe and environmentally compliant manner, the amount and form of tritium in the used nuclear fuel (UNF) must be understood and quantified.To gain a better understanding of how tritium in cladding will behave during processing, scoping tests are being performed to determine the tritium content in the cladding pre- and post-tritium pretreatment. A sample of Surry-2 pressurized water reactor (PWR) cladding was heated to 1100–1200°C to oxidize the zirconium and release all of the tritium in the cladding sample. The tritium content was measured to be ~240 µCi/g. Cladding samples were heated to 500ºC, which is within the temperature range (480 - 600ºC) expected for standard air tritium pretreatment systems, and to a slightly higher temperature (700ºC) to determine the impact of tritium pretreatment on tritium release from the cladding. Heating at 500°C for 24 hr removes ~0.2% of the tritium from the cladding, and heating at 700°C for 24 hr removes ~9%. Thus, a significant fraction of the tritium remains bound in the cladding and must be considered in operations involving cladding recycle.

  10. Facility Status

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Chinese Academy of Sciences, Hefei, Anhui, P.R. China The Engineering Design of ARC: A Compact, High Field, Fusion Nuclear Science Facility and Demonstration Power Plant B. N....

  11. Facility Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2002-05-20

    To establish facility safety requirements for the Department of Energy, including National Nuclear Security Administration. Cancels DOE O 420.1. Canceled by DOE O 420.1B.

  12. Facility Name Facility Name Facility FacilityType Owner Developer...

    Open Energy Info (EERE)

    FacilityStatus Coordinates D Metals D Metals D Metals Definition Small Scale Wind Valley City OH MW Northern Power Systems In Service AB Tehachapi Wind Farm AB Tehachapi...

  13. Facility Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2000-11-20

    The objective of this Order is to establish facility safety requirements related to: nuclear safety design, criticality safety, fire protection and natural phenomena hazards mitigation. The Order has Change 1 dated 11-16-95, Change 2 dated 10-24-96, and the latest Change 3 dated 11-22-00 incorporated. The latest change satisfies a commitment made to the Defense Nuclear Facilities Safety Board (DNFSB) in response to DNFSB recommendation 97-2, Criticality Safety.

  14. Facility Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2005-12-22

    The order establishes facility and programmatic safety requirements for nuclear and explosives safety design criteria, fire protection, criticality safety, natural phenomena hazards (NPH) mitigation, and the System Engineer Program.Chg 1 incorporates the use of DOE-STD-1189-2008, Integration of Safety into the Design Process, mandatory for Hazard Category 1, 2 and 3 nuclear facilities. Cancels DOE O 420.1A.

  15. Facility Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2013-06-21

    DOE-STD-1104 contains the Department's method and criteria for reviewing and approving nuclear facility's documented safety analysis (DSA). This review and approval formally document the basis for DOE, concluding that a facility can be operated safely in a manner that adequately protects workers, the public, and the environment. Therefore, it is appropriate to formally require implementation of the review methodology and criteria contained in DOE-STD-1104.

  16. Laser-induced synthesis and decay of Tritium under exposure of solid targets in heavy water

    E-Print Network [OSTI]

    E. V. Barmina; P. G. Kuzmin; S. F. Timashev; G. A. Shafeev

    2013-06-03

    The processes of laser-assisted synthesis of Tritium nuclei and their laser-induced decay in cold plasma in the vicinity of solid targets (Au, Ti, Se, etc.) immersed into heavy water are experimentally realized at peak laser intensity of 10E10-10E13 Watts per square centimeter. Initial stages of Tritium synthesis and their laser-induced beta-decay are interpreted on the basis of non-elastic interaction of plasma electrons having kinetic energy of 5-10 eV with nuclei of Deuterium and Tritium, respectively.

  17. TRITIUM PERMEATION AND TRANSPORT IN THE GASOLINE PRODUCTION SYSTEM COUPLED WITH HIGH TEMPERATURE GAS-COOLED REACTORS (HTGRS)

    SciTech Connect (OSTI)

    Chang H. Oh; Eung S. Kim; Mike Patterson

    2011-05-01

    This paper describes scoping analyses on tritium behaviors in the HTGR-integrated gasoline production system, which is based on a methanol-to-gasoline (MTG) plant. In this system, the HTGR transfers heat and electricity to the MTG system. This system was analyzed using the TPAC code, which was recently developed by Idaho National Laboratory. The global sensitivity analyses were performed to understand and characterize tritium behaviors in the coupled HTGR/MTG system. This Monte Carlo based random sampling method was used to evaluate maximum 17,408 numbers of samples with different input values. According to the analyses, the average tritium concentration in the product gasoline is about 3.05×10-3 Bq/cm3, and 62 % cases are within the tritium effluent limit (= 3.7x10-3 Bq/cm3[STP]). About 0.19% of released tritium is finally transported from the core to the gasoline product through permeations. This study also identified that the following four parameters are important concerning tritium behaviors in the HTGR/MTG system: (1) tritium source, (2) wall thickness of process heat exchanger, (3) operating temperature, and (4) tritium permeation coefficient of process heat exchanger. These four parameters contribute about 95 % of the total output uncertainties. This study strongly recommends focusing our future research on these four parameters to improve modeling accuracy and to mitigate tritium permeation into the gasol ine product. If the permeation barrier is included in the future study, the tritium concentration will be significantly reduced.

  18. Recovery of tritium dissolved in sodium at the steam generator of fast breeder reactor

    SciTech Connect (OSTI)

    Oya, Y.; Oda, T.; Tanaka, S.; Okuno, K.

    2008-07-15

    The tritium recovery technique in steam generators for fast breeder reactors using the double pipe concept was proposed. The experimental system for developing an effective tritium recovery technique was developed and tritium recovery experiments using Ar gas or Ar gas with 10-10000 ppm oxygen gas were performed using D{sub 2} gas instead of tritium gas. It was found that deuterium permeation through two membranes decreased by installing the double pipe concept with Ar gas. By introducing Ar gas with 10000 ppm oxygen gas, the concentration of deuterium permeation through two membranes decreased by more than 1/200, compared with the one pipe concept, indicating that most of the deuterium was scavenged by Ar gas or reacted with oxygen to form a hydroxide. However, most of the hydroxide was trapped at the surface of the membranes because of the short duration of the experiment. (authors)

  19. Is Tritium Over-regulated, Part 2: Should the TFG Support Higher...

    Office of Environmental Management (EM)

    (all oxide) RR Respiration Rate 3.33 E-4 m3sec (during plume passage) CDSE Cloud Shine Dose Equivalent 0 for tritium Q Threshold Value 31.2 for (RF1.0) or...

  20. Experiments on a ceramic electrolysis cell and a palladium diffuser at the tritium systems test assembly

    SciTech Connect (OSTI)

    Konishi, Satoshi; Yoshida, Hiroshi; Ohno, Hideo; Naruse, Yuji; Coffin, D.O.; Walthers, C.R.; Binning, K.E.

    1985-01-01

    A ceramic electrolysis cell and a palladium diffuser are developed in Japan and is tested with tritium in Tritium Systems Test Assembly (TSTA) of the Los Alamos National Laboratory, in order to confirm the feasibility as possible upgrades for the fuel cleanup system (PCU). The ceramic electrolysis cell made of stabilized zirconia was operated at 630/sup 0/C for an extended period with a mixture of 3% T/sub 2/O in He carrier gas in the circulation system with oxidizing catalyst bed. The palladium diffuser was tested with circulated pure tritium gas at 280/sup 0/C to verify the compatibility of the alloy with tritium, since the /sup 3/He produced in the metal could cause a degradation. The isotopic effects were also measured for both devices.

  1. Determination of the deuterium-tritium branching ratio based on inertial confinement fusion implosions

    E-Print Network [OSTI]

    Rosenberg, Michael Jonathan

    The deuterium-tritium (D-T) ?-to-neutron branching ratio [[superscript 3]H(d,?)[superscript 5]He/[superscript 3]H(d,n)[superscript 4]He] was determined under inertial confinement fusion (ICF) conditions, where the ...

  2. PPPL3157 Preprint Date: March 1996, UC421, 423, 426 Investigations of the Tritium Recycling

    E-Print Network [OSTI]

    reaction rate (product of cross section and relative velocity averaged over relative velocities). From Eqs). Because deuterium and tritium have the same Coulomb barrier, the mean reaction rates have nearly the same

  3. Design and Fabrication of In-Reactor Experiment to Measure Tritium...

    Office of Environmental Management (EM)

    Design and Fabrication of In-Reactor Experiment to Measure Tritium Release and Speciation from LiAlO2 and LiAlO2Zr Cermets Design and Fabrication of In-Reactor Experiment to...

  4. Feasibility of recoil enhanced tritium release from fusion blankets containing solid lithium compounds 

    E-Print Network [OSTI]

    Palmrose, Donald Edwin

    1986-01-01

    FEASIBILITY OF RECOIL ENHANCED TRITIUM RELEASE FROM FUSION BLANXETS CONTAINING SOLID LITHIUM COMPOUNDS A Thesis by DONALD EDWIN PALMROSE Submitted to the Graduate College of Texas A&M University in partial fulfillment of the requirements... for the degree of MASTER OF SCIENCE May 1986 Major Subject: Nuclear Engineering 1986 DONALD EDIJIi4 PAL;lROSE ALL RIGHTS RESERVED FEASIBILITY OF RECOIL ENHANCED TRITIUM RELEASE FROM FUSION BLANKETS CONTAINING SOLID LITHIUM COMPOUNDS A Thesis...

  5. Shock-tuned cryogenic-deuterium-tritium implosion performance on Omega

    SciTech Connect (OSTI)

    Sangster, T. C.; Goncharov, V. N.; Betti, R.; Boehly, T. R.; Collins, T. J. B.; Craxton, R. S.; Delettrez, J. A.; Edgell, D. H.; Epstein, R.; Glebov, Y. Yu.; Harding, D. R.; Hu, S. X.; Igumenschev, I. V.; Knauer, J. P.; Loucks, S. J.; Marozas, J. A.; Marshall, F. J.; McCrory, R. L.; McKenty, P. W.; Meyerhofer, D. D. [Laboratory for Laser Energetics, University of Rochester, 250 East River Road, Rochester, New York 14623 (United States)

    2010-05-15

    Cryogenic-deuterium-tritium (DT) target compression experiments with low-adiabat (alpha), multiple-shock drive pulses have been performed on the Omega Laser Facility [T. R. Boehly, D. L. Brown, R. S. Craxton et al., Opt. Commun. 133, 495 (1997)] to demonstrate hydrodynamic-equivalent ignition performance. The multiple-shock drive pulse facilitates experimental shock tuning using an established cone-in-shell target platform [T. R. Boehly, R. Betti, T. R. Boehly et al., Phys. Plasmas 16, 056301 (2009)]. These shock-tuned drive pulses have been used to implode cryogenic-DT targets with peak implosion velocities of 3x10{sup 7} cm/s at peak drive intensities of 8x10{sup 14} W/cm{sup 2}. During a recent series of alphaapprox2 implosions, one of the two necessary conditions for initiating a thermonuclear burn wave in a DT plasma was achieved: an areal density of approximately 300 mg/cm{sup 2} was inferred using the magnetic recoil spectrometer [J. A. Frenje, C. K. Li, F. H. Seguin et al., Phys. Plasmas 16, 042704 (2009)]. The other condition--a burn-averaged ion temperature {sub n} of 8-10 keV--cannot be achieved on Omega because of the limited laser energy; the kinetic energy of the imploding shell is insufficient to heat the plasma to these temperatures. A {sub n} of approximately 3.4 keV would be required to demonstrate ignition hydrodynamic equivalence [Betti et al., Phys. Plasmas17, 058102 (2010)]. The {sub n} reached during the recent series of alphaapprox2 implosions was approximately 2 keV, limited primarily by laser-drive and target nonuniformities. Work is underway to improve drive and target symmetry for future experiments.

  6. Tritium trapping in silicon carbide in contact with solid breeder under high flux isotope reactor irradiation

    SciTech Connect (OSTI)

    H. Katsui; Y. Katoh; A. Hasegawa; M. Shimada; Y. Hatano; T. Hinoki; S. Nogami; T. Tanaka; S. Nagata; T. Shikama

    2013-11-01

    The trapping of tritium in silicon carbide (SiC) injected from ceramic breeding materials was examined via tritium measurements using imaging plate (IP) techniques. Monolithic SiC in contact with ternary lithium oxide (lithium titanate and lithium aluminate) as a ceramic breeder was irradiated in the High Flux Isotope Reactor (HFIR) in Oak Ridge, Tennessee, USA. The distribution of photo-stimulated luminescence (PSL) of tritium in SiC was successfully obtained, which separated the contribution of 14C ß-rays to the PSL. The tritium incident from ceramic breeders was retained in the vicinity of the SiC surface even after irradiation at 1073 K over the duration of ~3000 h, while trapping of tritium was not observed in the bulk region. The PSL intensity near the SiC surface in contact with lithium titanate was higher than that obtained with lithium aluminate. The amount of the incident tritium and/or the formation of a Li2SiO3 phase on SiC due to the reaction with lithium aluminate under irradiation likely were responsible for this observation.

  7. Evolution of tritium from deuterided palladium subject to high electrical currents

    SciTech Connect (OSTI)

    Claytor, T.N.; Tuggle, D.G.; Taylor, S.F.

    1992-01-01

    An increase in the tritium level was detected in deuterium when various configurations of palladium foil or powder and silicon wafers or powder were subject to a high pulsed current. The deuterium, at one atmosphere pressure, and was circulated in a sealed loop containing the cell and an ionization chamber to measure the tritium increase as a function of time. Over 4800 hours of data, spanning 10 cells (including deuterium and hydrogen controls), were collected with this system. Average tritium production has varied from 0.02 to 0.2 nCi/h. Due to experimental constraints we have not been able to measure neutron output with these cells while simultaneously measuring the tritium increase. The question of tritium contamination in the palladium has been primarily resolved by the development of techniques that allow the palladium powder or foil to be reused. Various methods for increasing the tritium production, such as, increased current density, surface modifiers, and higher deuterium loading, will be discussed. 8 refs, 5 figs.

  8. Evolution of tritium from deuterided palladium subject to high electrical currents

    SciTech Connect (OSTI)

    Claytor, T.N.; Tuggle, D.G.; Taylor, S.F.

    1992-12-31

    An increase in the tritium level was detected in deuterium when various configurations of palladium foil or powder and silicon wafers or powder were subject to a high pulsed current. The deuterium, at one atmosphere pressure, and was circulated in a sealed loop containing the cell and an ionization chamber to measure the tritium increase as a function of time. Over 4800 hours of data, spanning 10 cells (including deuterium and hydrogen controls), were collected with this system. Average tritium production has varied from 0.02 to 0.2 nCi/h. Due to experimental constraints we have not been able to measure neutron output with these cells while simultaneously measuring the tritium increase. The question of tritium contamination in the palladium has been primarily resolved by the development of techniques that allow the palladium powder or foil to be reused. Various methods for increasing the tritium production, such as, increased current density, surface modifiers, and higher deuterium loading, will be discussed. 8 refs, 5 figs.

  9. A Fusion Development Facility on the Critical Path to Fusion Energy

    SciTech Connect (OSTI)

    Chan, V. S.; Stambaugh, R

    2011-01-01

    A fusion development facility (FDF) based on the tokamak approach with normal conducting magnetic field coils is presented. FDF is envisioned as a facility with the dual objective of carrying forward advanced tokamak (AT) physics and enabling the development of fusion energy applications. AT physics enables the design of a compact steady-state machine of moderate gain that can provide the neutron fluence required for FDF's nuclear science development objective. A compact device offers a uniquely viable path for research and development in closing the fusion fuel cycle because of the demand to consume only a moderate quantity of the limited supply of tritium fuel before the technology is in hand for breeding tritium.

  10. A fusion development facility on the critical path to fusion energy

    SciTech Connect (OSTI)

    Chan, Dr. Vincent; Canik, John; Peng, Yueng Kay Martin

    2011-01-01

    A fusion development facility (FDF) based on the tokamak approach with normal conducting magnetic field coils is presented. FDF is envisioned as a facility with the dual objective of carrying forward advanced tokamak (AT) physics and enabling the development of fusion energy applications. AT physics enables the design of a compact steady-state machine of moderate gain that can provide the neutron fluence required for FDF s nuclear science development objective. A compact device offers a uniquely viable path for research and development in closing the fusion fuel cycle because of the demand to consume only a moderate quantity of the limited supply of tritium fuel before the technology is in hand for breeding tritium.

  11. Facility Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1995-10-13

    Establishes facility safety requirements related to: nuclear safety design, criticality safety, fire protection and natural phenomena hazards mitigation. Cancels DOE 5480.7A, DOE 5480.24, DOE 5480.28 and Division 13 of DOE 6430.1A. Canceled by DOE O 420.1A.

  12. Facility Safety

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2012-12-04

    The Order establishes facility and programmatic safety requirements for DOE and NNSA for nuclear safety design criteria, fire protection, criticality safety, natural phenomena hazards (NPH) mitigation, and System Engineer Program. Cancels DOE O 420.1B, DOE G 420.1-2 and DOE G 420.1-3.

  13. Facility Operations and Maintenance Facilities Management

    E-Print Network [OSTI]

    Capogna, Luca

    Facility Operations and Maintenance Facilities Management D101 Facilities Management R -575/affirmative action institution. 354 3 373 4 373A,B,C,D 4 Alm8/31/12 #12;Facility Operations and Maintenance, B 5 1409 5 1403 5 1403 A, B 4 1408 3 1408 A,B,C 3 1610 3 #12;Facility Operations and Maintenance

  14. Secondary Startup Neutron Sources as a Source of Tritium in a Pressurized Water Reactor (PWR) Reactor Coolant System (RCS)

    SciTech Connect (OSTI)

    Shaver, Mark W.; Lanning, Donald D.

    2010-02-01

    The hypothesis of this paper is that the Zircaloy clad fuel source is minimal and that secondary startup neutron sources are the significant contributors of the tritium in the RCS that was previously assigned to release from fuel. Currently there are large uncertainties in the attribution of tritium in a Pressurized Water Reactor (PWR) Reactor Coolant System (RCS). The measured amount of tritium in the coolant cannot be separated out empirically into its individual sources. Therefore, to quantify individual contributors, all sources of tritium in the RCS of a PWR must be understood theoretically and verified by the sum of the individual components equaling the measured values.

  15. Toroidal Alfvén Eigenmodes in TFTR Deuterium-Tritium Plasmas

    SciTech Connect (OSTI)

    G.Y. Fu; H. Berk; R. Nazikian; S.H. Batha; Z. Chang; et al

    1998-01-01

    Purely alpha-particle-driven Toroidal Alfvén Eigenmodes (TAEs) with toroidal mode numbers n=1-6 have been observed in Deuterium-Tritium (D-T) plasmas on the Tokamak Fusion Test Reactor [D.J. Grove and D.M. Meade, Nucl. Fusion 25, 1167 (1985)]. The appearance of mode activity following termination of neutral beam injection in plasmas with q(0)>1 is generally consistent with theoretical predictions of TAE stability [G.Y. Fu et al., Phys. Plasmas 3, 4036 (1996]. Internal reflectometer measurements of TAE activity is compared with theoretical calculations of the radial mode structure. Core localization of the modes to the region of reduced central magnetic shear is confirmed, however the mode structure can deviate significantly from theoretical estimates. The peak measured TAE amplitude of delta n/n~10(superscript -4) at r/a~0.3-0.4 corresponds to delta B/B~10-5, while dB/B~10(superscript -8) is measured at the plasma edge. Enhanced alpha particle loss associated with TAE activity has not been observed.

  16. Compression of a spherically symmetric deuterium-tritium plasma liner onto a magnetized deuterium-tritium target

    SciTech Connect (OSTI)

    Santarius, J. F.

    2012-07-15

    Converging plasma jets may be able to reach the regime of high energy density plasmas (HEDP). The successful application of plasma jets to magneto-inertial fusion (MIF) would heat the plasma by fusion products and should increase the plasma energy density. This paper reports the results of using the University of Wisconsin's 1-D Lagrangian, radiation-hydrodynamics, fusion code BUCKY to investigate two MIF converging plasma jet test cases originally analyzed by Samulyak et al.[Physics of Plasmas 17, 092702 (2010)]. In these cases, 15 cm or 5 cm radially thick deuterium-tritium (DT) plasma jets merge at 60 cm from the origin and converge radially onto a DT target magnetized to 2 T and of radius 5 cm. The BUCKY calculations reported here model these cases, starting from the time of initial contact of the jets and target. Compared to the one-temperature Samulyak et al. calculations, the one-temperature BUCKY results show similar behavior, except that the plasma radius remains about twice as long near maximum compression. One-temperature and two-temperature BUCKY results differ, reflecting the sensitivity of the calculations to timing and plasma parameter details, with the two-temperature case giving a more sustained compression.

  17. Investigation of the potential impacts from tritium soil contamination in the CP-5 yard.

    SciTech Connect (OSTI)

    Hysong, R. J.

    1998-12-21

    Based on a review of available data, significant contributions to low-level tritium soil contamination in the CP-5 yard have been made by airborne tritium fallout and rainout from the CP-5 ventilation system stack. Based on the distribution of tritium in the yard, it is also likely that leaks in secondary system piping which lead to the cooling towers were a significant contributor to tritium in CP-5 yard subsurface soil. Based on the foregoing analysis, low-level tritium contamination will not prohibit the release of the yard for unrestricted use in the future. Worst case dose estimates based on very conservative assumptions indicate that a 25 rmem annual effective dose equivalent limit will not be exceeded under the most restrictive residential-use family farm scenario. Given the impermeable nature of the glacial till under CP-5, low-level concentrations of tritium may be occasionally detected in the deep well (3300 12D), but the peak concentration will not approach the levels calculated by RESRAD; however, continued monitoring of the deep well is recommended. To ensure that all sources of potential tritium release have been removed from the CP-5 complex, removal of tritiated water from each rod-out hole and an evaluation of the physical integrity of the rod-out holes is recommended. This will also allow for an evaluation of tritium concentrations in shallow groundwater under CP-5 by sampling groundwater that is currently being forced into the drain tile system. Additional surface and subsurface soil sampling and analysis will be required to determine the final release status of soils around the Building 330 complex relative to elevated concentrations of CS-137, CO-60,Co-57, and Eu-152 identified during the 1993 IT Corporation characterization. The potential radiological impact from isolated elevations of the latter radionuclides is relatively low and can be evaluated as part of the final status survey of outdoor areas surrounding the Building 330 complex. In summary, the following activities are recommended: Remove tritiated water from each rod-out hole; Monitor rod-out hole tritium concentrations as they fill up with shallow groundwater; Continue groundwater monitoring and Perform surface and subsurface soil sampling around the CP-5 complex as part of the final status survey.

  18. Science &Technology Facilities Council

    E-Print Network [OSTI]

    Science &Technology Facilities Council Science &Technology Facilities Council Science and Technology Facilities Council Annual Report and Accounts 2011-2012 Science and Technology Facilities Council Laboratory, Cheshire; UK Astronomy Technology Centre, Edinburgh; Chilbolton Observatory, Hampshire; Isaac

  19. Export Control Requirements for Tritium Processing Design and R&D

    SciTech Connect (OSTI)

    Hollis, William Kirk; Maynard, Sarah-Jane Wadsworth

    2015-10-30

    This document will address requirements of export control associated with tritium plant design and processes. Los Alamos National Laboratory has been working in the area of tritium plant system design and research and development (R&D) since the early 1970’s at the Tritium Systems Test Assembly (TSTA). This work has continued to the current date with projects associated with the ITER project and other Office of Science Fusion Energy Science (OS-FES) funded programs. ITER is currently the highest funding area for the DOE OS-FES. Although export control issues have been integrated into these projects in the past a general guidance document has not been available for reference in this area. To address concerns with currently funded tritium plant programs and assist future projects for FES, this document will identify the key reference documents and specific sections within related to tritium research. Guidance as to the application of these sections will be discussed with specific detail to publications and work with foreign nationals.

  20. The effect of water on tritium release behavior from solid breeder candidates

    SciTech Connect (OSTI)

    Suematsu, K.; Nishikawa, M.; Fukada, S.; Kinjyo, T.; Koyama, T.; Yamashita, N. [Graduate School of Engineering Science, Kyushu Univ., Fukuoka, 812-8581 (Japan)

    2008-07-15

    The authors have made a tritium release model to represent the release behavior of bred tritium from solid breeder materials using a series of studies. It has been observed that a large amount of adsorbed water and water produced by water formation reaction are released to the purge gas even though dry purge gas with hydrogen is introduced to solid breeder materials. According to our tritium release model, the presence of water in the purge gas and surface water on the material has a large effect on the tritium release behavior. In this study, the authors quantified the amount of adsorbed water and the capacity of the water formation reaction for various solid breeder materials (Li{sub 2}TiO{sub 3}, Li{sub 4}SiO{sub 4}, Li{sub 2}ZrO{sub 3}, LiAlO{sub 2}). The effect of surface water on the chemical form of tritium released from the LiAlO{sub 2} blanket is also discussed in this study. (authors)

  1. Tritium production from a low voltage deuterium discharge on palladium and other metals

    SciTech Connect (OSTI)

    Claytor, T.N.; Jackson, D.D.; Tuggle, D.G.

    1995-09-01

    Over the past year the authors have been able to demonstrate that a plasma loading method produces an exciting and unexpected amount of tritium from small palladium wires. In contrast to electrochemical hydrogen or deuterium loading of palladium, this method yields a reproducible tritium generation rate when various electrical and physical conditions are met. Small diameter wires (100--250 microns) have been used with gas pressures above 200 torr at voltages and currents of about 2,000 V at 3--5 A. By carefully controlling the sputtering rate of the wire, runs have been extended to hundreds of hours allowing a significant amount (> 10`s nCi) of tritium to accumulate. they show tritium generation rates for deuterium-palladium foreground runs that are up to 25 times larger than hydrogen-palladium control experiments using materials from the same batch. They illustrate the difference between batches of annealed palladium and as received palladium from several batches as well as the effect of other metals (Pt, Ni, Nb, Zr, V, W, Hf) to demonstrate that the tritium generation rate can vary greatly from batch to batch.

  2. Science Facilities

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of NaturalDukeWakefieldSulfateSciTechtail.Theory ofDidDevelopment Top LDRDUniversitySchedules PrintNIF About BlogFacilities

  3. Mobile Facility

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of NaturalDukeWakefieldSulfateSciTechtail.Theory ofDid you notHeatMaRIEdioxide capture |GE PutsgovSitesMobile Facility AMF

  4. Validation of the TIARA code to tritium inventory data

    SciTech Connect (OSTI)

    Billone, M.C.

    1994-03-01

    The TIARA code has been developed to predict tritium inventory in Li{sub 2}O breeder ceramic and to predict purge exit flow rate and composition. Inventory predictions are based on models for bulk diffusion, surface desorption, solubility and precipitation. Parameters for these models are determined from the results of laboratory annealing studies on unirradiated and irradiated Li{sub 2}O. Inventory data from in-reactor purge flow tests are used for model improvement, fine-tuning of model parameters and validation. In this current work, the inventory measurement near the purge inlet from the BEATRIX-II thin-ring sample is used to fine tune the surface desorption model parameters for T > 470{degrees}C, and the inventory measurement near the midplane from VOM-15H is used to fine tune the moisture solubility model parameters. predictions are then validated to the remaining inventory data from EXOTIC-2 (1 point), SIBELIUS (3 axial points), VOM-15H (2 axial points), CRITIC-1 (4 axial points), BEATRIX-II thin ring (3 axial points) and BEATRIX-II thick pellet (5 radial points). Thus. of the 20 data points, two we re used for fine tuning model parameters and 18 were used for validation. The inventory data span the range of 0.05--1.44 wppm with an average of 0.48 wppm. The data pertain to samples whose end-of-life temperatures were in the range of 490--1000{degrees}C. On the average, the TIARA predictions agree quite well with the data (< 0.02 wppm difference). However, the root-mean-square deviation is 0.44 wppm, mostly due to over-predictions for the SIBELIUS samples and the higher-temperature radial samples from the BEATRIX-11 thick-pellet.

  5. Design and Fabrication of In-Reactor Experiment to Measure Tritium Release and Speciation from LiAlO2 and LiAlO2/Zr Cermets

    Broader source: Energy.gov [DOE]

    Presentation from the 32nd Tritium Focus Group Meeting held in Germantown, Maryland on April 23-25, 2013.

  6. Test Results For Physical Separation Of Tritium From Noble Gases And It’s Implications For Sensitivity And Accuracy In Air And Stack Monitoring

    Broader source: Energy.gov [DOE]

    Presentation from the 32nd Tritium Focus Group Meeting held in Germantown, Maryland on April 23-25, 2013.

  7. ANALYSIS OF TRITIUM/DEUTERIUM RETENTION AND PERMEATION IN FW/DIVERTOR INCLUDING GEOMETRIC AND TEMPERATURE OPERATING FEATURES

    E-Print Network [OSTI]

    Abdou, Mohamed

    Department, UCLA, Los Angeles, CA 90095, USA ying@fusion.ucla.edu Available data and mathematical materials were implemented in the commercial code COMSOL Multiphysics. The goal is to develop a CAD-based hydrogen /tritium species and chemical compositions. As the data have shown, tritium has a higher

  8. IMPACT OF PRESSURE EQUALIZATION SLOT IN FLOW CHANNEL INSERT ON TRITIUM TRANSPORT IN A DCLL-TYPE POLOIDAL DUCT

    E-Print Network [OSTI]

    Abdou, Mohamed

    and thus the permeation rate. To reduce the stress in the FCI structure material, a PES is utilized 90095, USA, zhjbook@gmail.com A SiC-based flow channel insert (FCI) is used as an electrical and thermal the tritium transfer behavior and loss rate. Therefore it is important to examine the tritium loss rate

  9. Fusion Technologies for Tritium-Suppressed D-D Fusion White Paper prepared for FESAC Materials Science Subcommittee

    E-Print Network [OSTI]

    1 Fusion Technologies for Tritium-Suppressed D-D Fusion White Paper prepared for FESAC Materials, Columbia University 2 Plasma Science and Fusion Center, MIT December 19, 2011 Summary The proposal for tritium-suppressed D-D fusion and the understanding of the turbulent pinch in magnetically confined plasma

  10. PPPL-3172 -Preprint Date: May 1996, UC-420, 424, 426 Measurements of tritium retention and removal on TFTR.

    E-Print Network [OSTI]

    Abstract Recent experiments on the Tokamak Fusion Test Reactor (TFTR) have afforded an opportunity techniques were successful in removing 8,000 Ci and restoring the tritium inventory to a level well below in graphite which could raise the in-vessel tritium inventory to unacceptable levels unless an effective

  11. PPPL3172 Preprint Date: May 1996, UC420, 424, 426 Measurements of tritium retention and removal on TFTR.

    E-Print Network [OSTI]

    Abstract Recent experiments on the Tokamak Fusion Test Reactor (TFTR) have afforded an opportunity techniques were successful in removing » 8,000 Ci and restoring the tritium inventory to a level well below in graphite which could raise the in­vessel tritium inventory to unacceptable levels unless an effective

  12. Steady-State Impurity Control, Heat Removal and Tritium Recovery by Moving-Belt Plasma-Facing Components

    E-Print Network [OSTI]

    Tillack, Mark

    1 Steady-State Impurity Control, Heat Removal and Tritium Recovery by Moving-Belt Plasma-Z getter materials, heat removal and tritium recovery. In order to minimize MHD effects as well as induced is the application of "Moving-Belt Plasma-Facing Components" for steady-state impurity gettering, heat removal

  13. Recent palladium membrane reactor development at the tritium systems test assembly

    SciTech Connect (OSTI)

    Scott, W.R.; Birdsell, S.A.; Wilhelm, R.C. [Los Alamos National Lab., NM (United States)

    1995-10-01

    The palladium membrane reactor (PMR) is being investigated as a means for recovering hydrogen isotopes (including tritium) from compounds such as water and methane. Previous work with protiated water and methane showed that this device can be used to obtain high hydrogen recovery efficiencies using a single processing pass and with essentially no waste production. With these successful proof-of-principle results completed, recent work has focused on PMR development. This included studies of various geometries and testing with tritium. The results, which are reported here, have led to a better understanding of the PMR and will lead to the ultimate goal of building a production PMR and putting it into practical tritium processing service. 3 refs., 5 figs., 1 tab.

  14. PUREX facility hazards assessment

    SciTech Connect (OSTI)

    Sutton, L.N.

    1994-09-23

    This report documents the hazards assessment for the Plutonium Uranium Extraction Plant (PUREX) located on the US Department of Energy (DOE) Hanford Site. Operation of PUREX is the responsibility of Westinghouse Hanford Company (WHC). This hazards assessment was conducted to provide the emergency planning technical basis for PUREX. DOE Order 5500.3A requires an emergency planning hazards assessment for each facility that has the potential to reach or exceed the lowest level emergency classification. In October of 1990, WHC was directed to place PUREX in standby. In December of 1992 the DOE Assistant Secretary for Environmental Restoration and Waste Management authorized the termination of PUREX and directed DOE-RL to proceed with shutdown planning and terminal clean out activities. Prior to this action, its mission was to reprocess irradiated fuels for the recovery of uranium and plutonium. The present mission is to establish a passively safe and environmentally secure configuration at the PUREX facility and to preserve that condition for 10 years. The ten year time frame represents the typical duration expended to define, authorize and initiate follow-on decommissioning and decontamination activities.

  15. Lithium Ceramic Blankets for Russian Fusion Reactors and Influence of Breeding Operation Mode on Parameters of Reactor Tritium Systems

    SciTech Connect (OSTI)

    Kapyshev, Victor K.; Chernetsov, Mikhail Yu.; Zhevotov, Sergej I.; Kersnovskij, Alexandr Yu.; Kolbasov, Boris N.; Kovalenko, Victor G.; Paltusov, Nikolaj P.; Sernyaev, Georgeij A.; Sterebkov, Juri S.; Zyryanov, Alexej P.

    2005-07-15

    Russian controlled fusion program supposes development of a DEMO reactor design and participation in ITER Project. A solid breeder blanket of DEMO contains a ceramic lithium orthosilicate breeder and a beryllium multiplier. Test modules of the blanket are developed within the scope of ITER activities. Experimental models of module tritium breeding zones (TBZ), materials and fabrication technology of the TBZ, tritium reactor systems to analyse and process gas released from lithium ceramics are being developed. Two models of tritium breeding and neutron multiplying elements of the TBZ have been designed, manufactured and tested in IVV-2M nuclear reactor. Initial results of the in-pile experiments and outcome of lithium ceramics irradiation in a water-graphite nuclear reactor are considered to be a data base for development of the test modules and initial requirements for DEMO tritium system design. Influence of the tritium release parameters and hydrogen concentration in a purge gas on parameters of reactor system are discussed.

  16. Fourier Transform Infrared Spectroscopic Analysis Of Plastic Capsule Materials Exposed To Deuterium-Tritium (DT) Gas

    SciTech Connect (OSTI)

    Schoonover, J R; Steckle, Jr., W P; Elliot, N; Ebey, P S; Nobile, A; Nikroo, A; Cook, R C; Letts, S A

    2005-06-16

    Planar samples of varying thicknesses of both CH and CD glow discharge polymer have been measured with Fourier transform infrared (FTIR) spectroscopy before and after exposure to deuterium-tritium (DT) gas at elevated temperature and pressure. Planar samples of polyimide films made from both hydrogenated and deuterated precursors have also been examined by FTIR before and after DT exposure. The post-exposure FTIR spectra demonstrated no measurable exchange of hydrogen with deuterium or tritium for either polymer. Evidence for oxidation of the glow discharge polymer due to atmospheric oxygen was the only chemical change indicated by the FTIR data.

  17. An analysis of tritium and fissile fuel exchange in fusion-fission systems 

    E-Print Network [OSTI]

    Rice, Brent Lee

    1987-01-01

    AN ANALYSIS OF TRITIUM AND FISSILE FUEL EXCHANGE IN FUSION-FISSION SYSTEMS A Thesis by BRENT LEE RICE Submitted to the Graduate College of Texas A&M University in partial fulfillment of the requirement for the degree of MASTER OF SCIENCE... August 1987 Major Subject: Nuclear Engineering AN ANALYSIS OF TRITIUM AND FISSILE FUEL EXCHANGE IN FUSION-FISSION SYSTEMS A Thesis by BRENT LEE RICE Approved as to style and content by: T. A. Parish (Cha ir of Committee) R. R. Hart (Member) W...

  18. Tritium in the World Trade Center September 11, 2001 Terrorist Attack: It's Possible Sources and Fate

    SciTech Connect (OSTI)

    Parekh, P; Semkow, T; Husain, L; Haines, D; Woznial, G; Williams, P; Hafner, R; Rabun, R

    2002-05-03

    Traces of tritiated water (HTO) were determined at World Trade Center (WTC) ground zero after the 9/11/01 terrorist attack. A method of ultralow-background liquid scintillation counting was used after distilling HTO from the samples. A water sample from the WTC sewer, collected on 9/13/01, contained 0.174{plus_minus}0.074 (2{sigma}) nCi/L of HTO. A split water sample, collected on 9/21/01 from the basement of WTC Building 6, contained 3.53{plus_minus}0.17 and 2.83{plus_minus}0.15 nCi/L, respectively. Several water and vegetation samples were analyzed from areas outside the ground zero, located in Manhattan, Brooklyn, Queens, and Kensico Reservoir. No HTO above the background was found in those samples. All these results are well below the levels of concern to human exposure. Several tritium radioluminescent (RL) devices were investigated as possible sources of the traces of tritium at ground zero. Tritium is used in self-luminescent emergency EXIT signs. No such signs were present inside the WTC buildings. However, it was determined that Boeing 767-222 aircraft operated by the United Airlines that hit WTC Tower 2 as well as Boeing 767-223ER operated by the American Airlines, that hit WTC Tower 1, had a combined 34.3 Ci of tritium at the time of impact. Other possible sources of tritium include dials and lights of fire and emergency equipment, sights and scopes in weaponry, as well as time devices equipped with tritium dials. It was determined that emergency equipment was not a likely source. However, WTC hosted several law-enforcement agencies such as ATF, CIA, US Secret Service and US Customs. The ATF office had two weapon vaults in WTC Building 6. Also 63 Police Officers, possibly carrying handguns with tritium sights, died in the attack. The weaponry containing tritium was therefore a likely and significant source of tritium. It is possible that some of the 2830 victims carried tritium watches, however this source appears to be less significant that the other two. The fate of tritium in the attack depended on its chemistry. Any tritium present in the vicinity of jet-fuel explosion or fire would convert to HTO. The molecular tritium is also known to quickly exchange with water adsorbed on surfaces at ambient temperatures. Therefore, the end product of reacted tritium was HTO. A part of it would disperse into the atmosphere and a part would remain on site. The dynamic aspect of HTO removal was investigated taking into a consideration water flow at ground zero. Most of ground zero is encircled by the Slurry Wall, 70 ft deep underground, called a Bathtub. Approximately three million gallons of water were hosed on site in the fire-fighting efforts, and 1 million gallons fell as rainwater, between 9/11 and 9/21 (the day of the reported measurement). The combined water percolated through the debris down to the bottom of the Bathtub dissolving and removing HTO with it. That water would meet and combine with the estimated 26 million gallons of water that leaked from the Hudson River as well as broken mains, during the same period of 10 days after the attack. The combined water was collecting in the PATH train tunnel and continuously being pumped out to prevent flooding. A %Box model of water flow was developed to describe the above scenario. Considering the uncertainty in the amount of tritium present from sources other than the aircraft, as well as the dynamic character of tritium removal from the site, it is feasible to provide only a qualitative picture of the fate and behavior of tritium at WTC with the limited experimental data available. If the time history of tritium concentration at WTC had been measured, this study could have been a tracer study of water flow at WTC possibly useful to civil engineering.

  19. Apparatus for hydrocarbon extraction

    DOE Patents [OSTI]

    Bohnert, George W.; Verhulst, Galen G.

    2013-03-19

    Systems and methods for hydrocarbon extraction from hydrocarbon-containing material. Such systems and methods relate to extracting hydrocarbon from hydrocarbon-containing material employing a non-aqueous extractant. Additionally, such systems and methods relate to recovering and reusing non-aqueous extractant employed for extracting hydrocarbon from hydrocarbon-containing material.

  20. Proceedings of the tenth annual DOE low-level waste management conference: Session 3: Disposal technology and facility development

    SciTech Connect (OSTI)

    Not Available

    1988-12-01

    This document contains ten papers on various aspects of low-level radioactive waste management. Topics include: design and construction of a facility; alternatives to shallow land burial; the fate of tritium and carbon 14 released to the environment; defense waste management; engineered sorbent barriers; remedial action status report; and the disposal of mixed waste in Texas. Individual papers were processed separately for the data base. (TEM)

  1. Evidence for Stratification of Deuterium-Tritium Fuel in Inertial Confinement Fusion Implosions

    E-Print Network [OSTI]

    Casey, Daniel Thomas

    Measurements of the D(d,p)T (dd) and T(t,2n)[superscript 4]He (tt) reaction yields have been compared with those of the D(t,n)[superscript 4]He (dt) reaction yield, using deuterium-tritium gas-filled inertial confinement ...

  2. Detecting non-relativistic cosmic neutrinos by capture on tritium: phenomenology and physics potential

    E-Print Network [OSTI]

    Andrew J. Long; Cecilia Lunardini; Eray Sabancilar

    2014-11-12

    We study the physics potential of the detection of the Cosmic Neutrino Background via neutrino capture on tritium, taking the proposed PTOLEMY experiment as a case study. With the projected energy resolution of $\\Delta \\sim$ 0.15 eV, the experiment will be sensitive to neutrino masses with degenerate spectrum, $m_1 \\simeq m_2 \\simeq m_3 = m_\

  3. PPPL-3458 PPPL-3458 Visual Tritium Imaging Of In-Vessel Surfaces

    E-Print Network [OSTI]

    on the U.S. Department of Energy's Princeton Plasma Physics Laboratory Publications and Reports web site and DOE Contractors can obtain copies of this report from: U.S. Department of Energy Office of Scientific Energy Research Institute, Tritium Engineering Laboratory, Tokai, Ibaraki 319-1195, Japan Abstract

  4. TRITIUM ANALYSIS OF A WATER-COOLED SOLID BREEDER BLANKET FOR ITER*

    E-Print Network [OSTI]

    Abdou, Mohamed

    .A. Abdou Mechanical,Aerospace and Nuclear Engineering Department University of California, Los Angeles Los at reduced power level. Key parameters affecting the kinetics of the tritium release and the inventory. The blanket uses beryllium for neutron multiplication and lithium-base ceramic such as oxide or orthosilicate

  5. F1-CN-64/AP2-17 DEUTERIUM-TRITIUM TFTR PLASMAS WITH HIGH

    E-Print Network [OSTI]

    Budny, Robert

    . The fusion power output, Pf, in high temperature tokamak plasmas heated by neutral beam injection scales by increasing the product of li Ip . Technical aspects of generating the high li plasma in previous experimentsF1-CN-64/AP2-17 DEUTERIUM-TRITIUM TFTR PLASMAS WITH HIGH INTERNAL INDUCTANCE* S. A. SABBAGH1 , E. D

  6. Behavior of Elemental Tritium in Atmospheric Diffusion in Inertial Fusion Reactors

    SciTech Connect (OSTI)

    Velarde, Marta; Perlado, Jose Manuel [Universidad Politecnica de Madrid (Spain)

    2003-05-15

    The evaluation of the radiological environmental impact of tritium emission to the atmosphere from inertial fusion energy (IFE) reactors has different chronological phases. In the release primary phase, the important factors are the boundary conditions: atmospheric and geometric grid from the point of emission. The second phase occurs when the tritium is deposited in the ground. This phase is important in order to account for the dosimetric effects of tritium, and it is a key factor in the chronic and collective doses of the population.The final internal irradiation dose is calculated as the addition of doses by ingestion, by inhalation of the primary plume, by absorption on the skin, and inhalation by reemission to the atmosphere.Each of the two chemical forms (HT and HTO) of tritium present in the environment from potential IFE reactor releases contributes in different ways to the most exposed individual and the committed effective dose equivalent (50-CEDE). The HTO presents a much larger percentage of the internal irradiation from inhalation and absorption through the skin than HT. However, in releases where HT represents 100%, the contributions to the total effective dose by ingestion and reemission are important.

  7. Examination of 80 deg. C desorption isotherms of tritium aged Pd/k and LANA.75

    SciTech Connect (OSTI)

    Staack, G. C.; Shanahan, K. L.; Walters, R. T.; Pilgrim, R. D.

    2008-07-15

    Metal hydrides, specifically Pd deposited on kieselguhr (Pd/k) and LaNi{sub 4.25}Al{sub 0.75} (LANA.75), have been used at the Savannah River Site for almost twenty years for hydrogen isotope separation and storage. Radiolytic decay of tritium to helium-3 in the metal matrix causes three classic changes in the performance of the hydride: the plateau pressure decreases, the plateau slope increases, and a heel forms, reducing the reversible capacity of the hydride. Deuterium and tritium isotherms were collected on the virgin materials, only tritium isotherms were collected at approximately 2 years, and both deuterium and tritium isotherms were collected at approximately 3.5 years of quiescent aging at 26 deg. C. Each sample was loaded to 0.5-0.6 T/M prior to each aging period. Points of interest include comparisons of each sample at different aging periods and isotope effects on aged hydride isotherms. Partial restoration of thermodynamic properties by sample cycling has been observed in LANA. 75, though not previously reported in Pd. The methods and results are presented. (authors)

  8. Modeling and analysis of time-dependent tritium transport in lithium oxide

    E-Print Network [OSTI]

    Raffray, A. René

    evaluation of the conditions for attaining self-suciency is necessary to de®ne the selection criteria for design concepts and the range of acceptable parameters [1]. Tritium behavior and transport in the blanket as steady state conditions. MISTRAL was Journal of Nuclear Materials 273 (1999) 79±94 www

  9. MANAGING BERYLLIUM IN NUCLEAR FACILITY APPLICATIONS

    SciTech Connect (OSTI)

    R. Rohe; T. N. Tranter

    2011-12-01

    Beryllium plays important roles in nuclear facilities. Its neutron multiplication capability and low atomic weight make it very useful as a reflector in fission reactors. Its low atomic number and high chemical affinity for oxygen have led to its consideration as a plasma-facing material in fusion reactors. In both applications, the beryllium and the impurities in it become activated by neutrons, transmuting them to radionuclides, some of which are long-lived and difficult to dispose of. Also, gas production, notably helium and tritium, results in swelling, embrittlement, and cracking, which means that the beryllium must be replaced periodically, especially in fission reactors where dimensional tolerances must be maintained. It has long been known that neutron activation of inherent iron and cobalt in the beryllium results in significant {sup 60}Co activity. In 2001, it was discovered that activation of naturally occurring contaminants in the beryllium creates sufficient {sup 14}C and {sup 94}Nb to render the irradiated beryllium 'Greater-Than-Class-C' for disposal in U.S. radioactive waste facilities. It was further found that there was sufficient uranium impurity in beryllium that had been used in fission reactors up to that time that the irradiated beryllium had become transuranic in character, making it even more difficult to dispose of. In this paper we review the extent of the disposal issue, processes that have been investigated or considered for improving the disposability of irradiated beryllium, and approaches for recycling.

  10. Simulation of background from low-level tritium and radon emanation in the KATRIN spectrometers

    SciTech Connect (OSTI)

    Leiber, B. [Institute for Nuclear Physics (IKP), Karlsruhe Institute of Technology (KIT), 76021 Karlsruhe (Germany)] [Institute for Nuclear Physics (IKP), Karlsruhe Institute of Technology (KIT), 76021 Karlsruhe (Germany); Collaboration: KATRIN Collaboration

    2013-08-08

    The KArlsruhe TRItium Neutrino (KATRIN) experiment is a large-scale experiment for the model independent determination of the mass of electron anti-neutrinos with a sensitivity of 200 meV/c{sup 2}. It investigates the kinematics of electrons from tritium beta decay close to the endpoint of the energy spectrum at 18.6 keV. To achieve a good signal to background ratio at the endpoint, a low background rate below 10{sup ?2} counts per second is required. The KATRIN setup thus consists of a high luminosity windowless gaseous tritium source (WGTS), a magnetic electron transport system with differential and cryogenic pumping for tritium retention, and electro-static retarding spectrometers (pre-spectrometer and main spectrometer) for energy analysis, followed by a segmented detector system for counting transmitted beta-electrons. A major source of background comes from magnetically trapped electrons in the main spectrometer (vacuum vessel: 1240 m{sup 3}, 10{sup ?11} mbar) produced by nuclear decays in the magnetic flux tube of the spectrometer. Major contributions are expected from short-lived radon isotopes and tritium. Primary electrons, originating from these decays, can be trapped for hours, until having lost almost all their energy through inelastic scattering on residual gas particles. Depending on the initial energy of the primary electron, up to hundreds of low energetic secondary electrons can be produced. Leaving the spectrometer, these electrons will contribute to the background rate. This contribution describes results from simulations for the various background sources. Decays of {sup 219}Rn, emanating from the main vacuum pump, and tritium from the WGTS that reaches the spectrometers are expected to account for most of the background. As a result of the radon alpha decay, electrons are emitted through various processes, such as shake-off, internal conversion and the Auger deexcitations. The corresponding simulations were done using the KASSIOPEIA framework, which has been developed for the KATRIN experiment for low-energy electron tracking, field calculation and detector simulation. The results of the simulations have been used to optimize the design parameters of the vacuum system with regard to radon emanation and tritium pumping, in order to reach the stringent requirements of the neutrino mass measurement.

  11. Control technology for radioactive emissions to the atmosphere at US Department of Energy facilities

    SciTech Connect (OSTI)

    Moore, E.B.

    1984-10-01

    The purpose of this report is to provide information to the US Environmental Protection agency (EPA) on existing technology for the control of radionuclide emissions into the air from US Department of Energy (DOE) facilities, and to provide EPA with information on possible additional control technologies that could be used to further reduce these emissions. Included in this report are generic discussions of emission control technologies for particulates, iodine, rare gases, and tritium. Also included are specific discussions of existing emission control technologies at 25 DOE facilities. Potential additional emission control technologies are discussed for 14 of these facilities. The facilities discussed were selected by EPA on the basis of preliminary radiation pathway analyses. 170 references, 131 figures, 104 tables.

  12. Nuclear Technology & Canadian Oil Sands: Integration of Nuclear Power with In-Situ Oil Extraction

    E-Print Network [OSTI]

    Nuclear Technology & Canadian Oil Sands: Integration of Nuclear Power with In-Situ Oil Extraction A for a Canadian oil sands extraction facility using Steam-Assisted Gravity Drainage (SAGD) technology. The energy to produce steam as well as electricity for the oil sands facility; and (3) using the reactor to produce

  13. Tritium Transport at the Rulison Site, a Nuclear-stimulated Low-permeability Natural Gas Reservoir

    SciTech Connect (OSTI)

    C. Cooper; M. Ye; J. Chapman

    2008-04-01

    The U.S. Department of Energy (DOE) and its predecessor agencies conducted a program in the 1960s and 1970s that evaluated technology for the nuclear stimulation of low-permeability natural gas reservoirs. The second project in the program, Project Rulison, was located in west-central Colorado. A 40-kiltoton nuclear device was detonated 2,568 m below the land surface in the Williams Fork Formation on September 10, 1969. The natural gas reservoirs in the Williams Fork Formation occur in low permeability, fractured sandstone lenses interbedded with shale. Radionuclides derived from residual fuel products, nuclear reactions, and activation products were generated as a result of the detonation. Most of the radionuclides are contained in a cooled, solidified melt glass phase created from vaporized and melted rock that re-condensed after the test. Of the mobile gas-phase radionuclides released, tritium ({sup 3}H or T) migration is of most concern. The other gas-phase radionuclides ({sup 85}Kr, {sup 14}C) were largely removed during production testing in 1969 and 1970 and are no longer present in appreciable amounts. Substantial tritium remained because it is part of the water molecule, which is present in both the gas and liquid (aqueous) phases. The objectives of this work are to calculate the nature and extent of tritium contamination in the subsurface from the Rulison test from the time of the test to present day (2007), and to evaluate tritium migration under natural-gas production conditions to a hypothetical gas production well in the most vulnerable location outside the DOE drilling restriction. The natural-gas production scenario involves a hypothetical production well located 258 m horizontally away from the detonation point, outside the edge of the current drilling exclusion area. The production interval in the hypothetical well is at the same elevation as the nuclear chimney created by the detonation, in order to evaluate the location most vulnerable to tritium migration.

  14. A search for neutrons and gamma rays associated with tritium production in deuterated metals

    SciTech Connect (OSTI)

    Wolf, K.L.; Lawson, D.R.; Packham, N.J.C.; Wass, J.C.

    1989-01-01

    Tritium activity has been measured in several Pd-Ni-D{sub 2}O electrolytic cells, as reported previously. At the present time 13 separate cells have shown tritium at 10{sup 2} to 10{sup 6} times the background level of the D{sub 2}O used in these experiments. The appearance of the activity in the electrolyte and in the gas phase occurs over a period of hours to a few days after remaining at or near the background level during 4--10 weeks of charging in 0.1 M LiOD, D{sub 2}O solution. The present paper deals with attempts to reproduce the tritium measurements and to establish the source, from either contamination or nuclear reaction. The sudden appearance of tritium activity in the cells requires the tritium to be loaded in a component prior to the beginning of cell operation in a contamination model. Release is assumed to be caused by deterioration of one of the materials used in the 0.1 M LiOD solution. In an extensive set of tests, no contamination has been found in the starting materials or in normal water blanks. Results for neutron and gamma-ray correlations have proved to be negative also. The limit set on the absence of 2.5 MeV neutrons for the t/n ration is 10{sup 7} from that expected in the d + d reaction, and 10{sup 3} for 14 MeV neutrons expected from the t + d secondary reaction. Similarly, Coulomb excitation gamma rays expected from the interaction of 3 MeV protons with Pd are found to be absent, which indicates that the d(d,p)t two-body reaction does not occur in the Pd electrode. 9 figs., 2 tabs.

  15. Guide to research facilities

    SciTech Connect (OSTI)

    Not Available

    1993-06-01

    This Guide provides information on facilities at US Department of Energy (DOE) and other government laboratories that focus on research and development of energy efficiency and renewable energy technologies. These laboratories have opened these facilities to outside users within the scientific community to encourage cooperation between the laboratories and the private sector. The Guide features two types of facilities: designated user facilities and other research facilities. Designated user facilities are one-of-a-kind DOE facilities that are staffed by personnel with unparalleled expertise and that contain sophisticated equipment. Other research facilities are facilities at DOE and other government laboratories that provide sophisticated equipment, testing areas, or processes that may not be available at private facilities. Each facility listing includes the name and phone number of someone you can call for more information.

  16. Tritium efflux from TFTR during a vacuum vessel vent1 D. Mueller, C.H. Skinner, W. Blanchard, A. Nagy

    E-Print Network [OSTI]

    Budny, Robert

    on the 1987 measurements of the outgassing rate of the tritium produced by D-D reactions, 34 Ci/day with a 4 VALVES Flow rate up to 2 x

  17. WBN-1 Cycle 10 TPBAR Tritium Release, Deduced From Analysis of RCS Data TTP-1-3046-00, Rev 0

    SciTech Connect (OSTI)

    Shaver, Mark W.; Niehus, Mark T.; Love, Edward F.

    2012-02-19

    This document contains the calculation of the TPBAR tritium release from the Mark 9.2 design TPBARs irradiated in WBN cycle 10. The calculation utilizes the generalized cycle analysis methodology given in TTP-1-3045 Rev. 0.

  18. Tritium production analysis and management strategies for a Fluoride-salt-cooled high-temperature test reactor (FHTR)

    E-Print Network [OSTI]

    Rodriguez, Judy N

    2013-01-01

    The Fluoride-salt-cooled High-temperature Test Reactor (FHTR) is a test reactor concept that aims to demonstrate the neutronics, thermal-hydraulics, materials, tritium management, and to address other reactor operational ...

  19. Radioactive Liquid Waste Treatment Facility Discharges in 2011

    SciTech Connect (OSTI)

    Del Signore, John C.

    2012-05-16

    This report documents radioactive discharges from the TA50 Radioactive Liquid Waste Treatment Facilities (RLWTF) during calendar 2011. During 2011, three pathways were available for the discharge of treated water to the environment: discharge as water through NPDES Outfall 051 into Mortandad Canyon, evaporation via the TA50 cooling towers, and evaporation using the newly-installed natural-gas effluent evaporator at TA50. Only one of these pathways was used; all treated water (3,352,890 liters) was fed to the effluent evaporator. The quality of treated water was established by collecting a weekly grab sample of water being fed to the effluent evaporator. Forty weekly samples were collected; each was analyzed for gross alpha, gross beta, and tritium. Weekly samples were also composited at the end of each month. These flow-weighted composite samples were then analyzed for 37 radioisotopes: nine alpha-emitting isotopes, 27 beta emitters, and tritium. These monthly analyses were used to estimate the radioactive content of treated water fed to the effluent evaporator. Table 1 summarizes this information. The concentrations and quantities of radioactivity in Table 1 are for treated water fed to the evaporator. Amounts of radioactivity discharged to the environment through the evaporator stack were likely smaller since only entrained materials would exit via the evaporator stack.

  20. EFFECTS OF TRITIUM GAS EXPOSURE ON THE GLASS TRANSITION TEMPERATURE OF EPDM ELASTOMER AND ON THE CONDUCTIVITY OF POLYANILINE

    SciTech Connect (OSTI)

    Clark, E; Marie Kane, M

    2008-12-12

    Four formulations of EPDM (ethylene-propylene diene monomer) elastomer were exposed to tritium gas initially at one atmosphere and ambient temperature for between three and four months in closed containers. Material properties that were characterized include density, volume, mass, appearance, flexibility, and dynamic mechanical properties. The glass transition temperature was determined by analysis of the dynamic mechanical property data per ASTM standards. EPDM samples released significant amounts of gas when exposed to tritium, and the glass transition temperature increased by about 3 C. during the exposure. Effects of ultraviolet and gamma irradiation on the surface electrical conductivity of two types of polyaniline films are also documented as complementary results to planned tritium exposures. Future work will determine the effects of tritium gas exposure on the electrical conductivity of polyaniline films, to demonstrate whether such films can be used as a sensor to detect tritium. Surface conductivity was significantly reduced by irradiation with both gamma rays and ultraviolet light. The results of the gamma and UV experiments will be correlated with the tritium exposure results.

  1. Evaluation of Hydrogen Isotope Exchange Methodology on Adsorbents for Tritium Removal

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Morgan, Gregg A.; Xiao, S. Xin

    2015-03-06

    The Savannah River National Laboratory has demonstrated a potential process that can be used to remove tritium from tritiated water using Pt-catalyzed molecular sieves. The process is an elemental isotope exchange process in which H2 (when flowed through the molecular sieves) will exchange with the adsorbed water, D2O, leaving H2O adsorbed on the molecular sieves. Various formulations of catalyzed molecular sieve material were prepared using two different techniques, Pt-implantation and Pt-ion exchange. This technology has been demonstrated for a protium (H) and deuterium (D) system, but can also be used for the removal of tritium from contaminated water (T2O, HTO,more »and DTO) using D2 (or H2)« less

  2. Oxidation of zirconium alloys in 2.5 kPa water vapor for tritium readiness.

    SciTech Connect (OSTI)

    Mills, Bernice E.

    2007-11-01

    A more reactive liner material is needed for use as liner and cruciform material in tritium producing burnable absorber rods (TPBAR) in commercial light water nuclear reactors (CLWR). The function of these components is to convert any water that is released from the Li-6 enriched lithium aluminate breeder material to oxide and hydrogen that can be gettered, thus minimizing the permeation of tritium into the reactor coolant. Fourteen zirconium alloys were exposed to 2.5 kPa water vapor in a helium stream at 300 C over a period of up to 35 days. Experimental alloys with aluminum, yttrium, vanadium, titanium, and scandium, some of which also included ternaries with nickel, were included along with a high nitrogen impurity alloy and the commercial alloy Zircaloy-2. They displayed a reactivity range of almost 500, with Zircaloy-2 being the least reactive.

  3. Titanium tritide radioisotope heat source development : palladium-coated titanium hydriding kinetics and tritium loading tests.

    SciTech Connect (OSTI)

    Van Blarigan, Peter; Shugard, Andrew D.; Walters, R. Tom

    2012-01-01

    We have found that a 180 nm palladium coating enables titanium to be loaded with hydrogen isotopes without the typical 400-500 C vacuum activation step. The hydriding kinetics of Pd coated Ti can be described by the Mintz-Bloch adherent film model, where the rate of hydrogen absorption is controlled by diffusion through an adherent metal-hydride layer. Hydriding rate constants of Pd coated and vacuum activated Ti were found to be very similar. In addition, deuterium/tritium loading experiments were done on stacks of Pd coated Ti foil in a representative-size radioisotope heat source vessel. The experiments demonstrated that such a vessel could be loaded completely, at temperatures below 300 C, in less than 10 hours, using existing department-of-energy tritium handling infrastructure.

  4. Fuel assembly for the production of tritium in light water reactors

    DOE Patents [OSTI]

    Cawley, W.E.; Trapp, T.J.

    1983-06-10

    A nuclear fuel assembly is described for producing tritium in a light water moderated reactor. The assembly consists of two intermeshing arrays of subassemblies. The first subassemblies comprise concentric annular elements of an outer containment tube, an annular target element, an annular fuel element, and an inner neutron spectrums shifting rod. The second subassemblies comprise an outer containment tube and an inner rod of either fuel, target, or neutron spectrum shifting neutral.

  5. Tritium: a model for low level long-term ionizing radiation exposure

    SciTech Connect (OSTI)

    Carsten, A.L.

    1984-01-01

    The somatic, cytogenetic and genetic effects of single and chronic tritiated water (HTO) ingestion in mice was investigated. This study serves not only as an evaluation of tritium toxicity (TRITOX) but due to its design involving long-term low concentration ingestion of HTO may serve as a model for low level long-term ionizing radiation exposure in general. Long-term studies involved animals maintained on HTO at concentrations of 0.3 ..mu..Ci/ml, 1.0 ..mu..Ci/ml, 3.0 ..mu..Ci/ml or depth dose equivalent chronic external exposures to /sup 137/Cs gamma rays. Maintenance on 3.0 ..mu..Ci/ml resulted in no effect on growth, life-time shortening or bone marrow cellularity, but did result in a reduction of bone marrow stem cells, an increase in DLM's in second generation animals maintained on this regimen and cytogenetic effects as indicated by increased sister chromatid exchanges (SCE's) in bone marrow cells, increased chromosome aberrations in the regenerating liver and an increase in micronuclei in red blood cells. Biochemical and microdosimetry studies showed that animals placed on the HTO regimen reached tritium equilibrium in the body water in approximately 17 to 21 days with a more gradual increase in bound tritium. When animals maintained for 180 days on 3.0 ..mu..Ci/ml HTO were placed on a tap water regimen, the tritium level in tissue dropped from the equilibrium value of 2.02 ..mu..Ci/ml before withdrawal to 0.001 ..mu..Ci/ml at 28 days. 18 references.

  6. Tritiation of aerogel matrices: T sub 2 O, tritiated organics and tritium exchange on aerogel surfaces

    SciTech Connect (OSTI)

    Ellefson, R.E.; Gill, J.T. (EG and G Mound Applied Technologies, Miamisburg, OH (USA)); Shepodd, T.J. (Sandia Labs., Livermore, CA (USA)); Leonard, L.E. (USDOE, Washington, DC (USA))

    1990-01-01

    Three methods for incorporation of tritium into the phoshor/aerogel matrix have been demonstrated: (1) adsorption of T{sub 2}O by the aerogel, (2) incorporation of tritiated organic into the pores of the aerogel and (3) isotopic exchange of tritium from T{sub 2} gas for the H residing on the surface of the aerogel. Adsorption of T{sub 2}O produces the brightest light (4.4 fL) to date but the tritium is loosely bound. Incorporation of tritiated organics into the pores of the aerogel produces less that theoretical luminance and intensity diminishes rapidly due to precipitation and darkening of the organic from radiation damage. Isotopic exchange produces a stable lamp by tritiating H sites on the surface of the aerogel. A lamp with stable luminance of 1.1 fL has been produced; a theoretical limit for a mono-layer coverage fo the aerogel surface is 2 to 3 fL. 7 refs., 4 figs., 2 tabs.

  7. CRAD, Facility Safety- Nuclear Facility Design

    Office of Energy Efficiency and Renewable Energy (EERE)

    A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) that can be used for assessment of a contractor's Nuclear Facility Design.

  8. CRAD, Facility Safety- Nuclear Facility Safety Basis

    Broader source: Energy.gov [DOE]

    A section of Appendix C to DOE G 226.1-2 "Federal Line Management Oversight of Department of Energy Nuclear Facilities." Consists of Criteria Review and Approach Documents (CRADs) that can be used for assessment of a contractor's Nuclear Facility Safety Basis.

  9. Catalytic extraction processing of contaminated scrap metal

    SciTech Connect (OSTI)

    Griffin, T.P.; Johnston, J.E.; Payea, B.M.; Zeitoon, B.M.

    1995-12-01

    Molten Metal Technology was awarded a contract to demonstrate the applicability of the Catalytic Extraction Process, a proprietary process that could be applied to US DOE`s inventory of low level mixed waste. This paper is a description of that technology, and included within this document are discussions of: (1) Program objectives, (2) Overall technology review, (3) Organic feed conversion to synthetic gas, (4) Metal, halogen, and transuranic recovery, (5) Demonstrations, (6) Design of the prototype facility, and (7) Results.

  10. Measurement of Energy Distribution of Deuterium-Tritium Fusion Alpha-particles and MeV Energy Knock-on Deuterons in JET Plasmas

    E-Print Network [OSTI]

    Measurement of Energy Distribution of Deuterium-Tritium Fusion Alpha-particles and MeV Energy Knock-on Deuterons in JET Plasmas

  11. Information extraction system

    DOE Patents [OSTI]

    Lemmond, Tracy D; Hanley, William G; Guensche, Joseph Wendell; Perry, Nathan C; Nitao, John J; Kidwell, Paul Brandon; Boakye, Kofi Agyeman; Glaser, Ron E; Prenger, Ryan James

    2014-05-13

    An information extraction system and methods of operating the system are provided. In particular, an information extraction system for performing meta-extraction of named entities of people, organizations, and locations as well as relationships and events from text documents are described herein.

  12. Waste Management facilities fault tree databank 1995 status report

    SciTech Connect (OSTI)

    Minnick, W.V.; Wellmaker, K.A.

    1995-08-16

    The Safety Information Management and Analysis Group (SIMA) of the Safety Engineering Department (SED) maintains compilations of incidents that have occurred in the Separations and Process Control, Waste Management, Fuel Fabrication, Tritium and SRTC facilities. This report records the status of the Waste Management (WM) Databank at the end of CY-1994. The WM Databank contains more than 35,000 entries ranging from minor equipment malfunctions to incidents with significant potential for injury or contamination of personnel. This report documents the status of the WM Databank including the availability, training, sources of data, search options, Quality Assurance, and usage to which these data have been applied. Periodic updates to this memorandum are planned as additional data or applications are acquired.

  13. Inertial Confinement Fusion and the National Ignition Facility (NIF)

    SciTech Connect (OSTI)

    Ross, P.

    2012-08-29

    Inertial confinement fusion (ICF) seeks to provide sustainable fusion energy by compressing frozen deuterium and tritium fuel to extremely high densities. The advantages of fusion vs. fission are discussed, including total energy per reaction and energy per nucleon. The Lawson Criterion, defining the requirements for ignition, is derived and explained. Different confinement methods and their implications are discussed. The feasibility of creating a power plant using ICF is analyzed using realistic and feasible numbers. The National Ignition Facility (NIF) at Lawrence Livermore National Laboratory is shown as a significant step forward toward making a fusion power plant based on ICF. NIF is the world’s largest laser, delivering 1.8 MJ of energy, with a peak power greater than 500 TW. NIF is actively striving toward the goal of fusion energy. Other uses for NIF are discussed.

  14. Polymer Exposure and Testing Facilities at the Savannah River Site

    Broader source: Energy.gov [DOE]

    Presentation from the 33rd Tritium Focus Group Meeting held in Aiken, South Carolina on April 22-24, 2014.

  15. Stockpile Stewardship and the National Ignition Facility

    SciTech Connect (OSTI)

    Moses, E

    2012-01-04

    The National Ignition Facility (NIF), the world's most energetic laser system, is operational at Lawrence Livermore National Laboratory (LLNL). Since the completion of the construction project in March 2009, NIF has completed nearly 150 target experiments for the National Ignition Campaign (NIC), High Energy Density Stewardship Science (HEDSS) in the areas of radiation transport, material dynamics at high pressure in the solid state, as well as fundamental science and other national security missions. NIF capabilities and infrastructure are in place to support all of its missions with over 50 X-ray, optical and nuclear diagnostic systems and the ability to shoot cryogenic targets and DT layered capsules. NIF is now qualified for use of tritium and other special materials as well as to perform high yield experiments and classified experiments. DT implosions with record indirect-drive neutron yield of 4.5 x 10{sup 14} neutrons have been achieved. A series of 43 experiments were successfully executed over a 27-day period, demonstrating the ability to perform precise experiments in new regimes of interest to HEDSS. This talk will provide an update of the progress on the NIF capabilities, NIC accomplishments, as well as HEDSS and fundamental science experimental results and an update of the experimental plans for the coming year.

  16. Facilities | Jefferson Lab

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    JLab Buildings Facilities Management & Logistics is responsible for performing or specifying performance of all Jefferson Lab facility maintenance. A D D I T I O N A L L I N K S:...

  17. Better building: LEEDing new facilities

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Better building: LEEDing new facilities Better building: LEEDing new facilities We're taking big steps on-site to create energy efficient facilities and improve infrastructure....

  18. Computing Facilities Orientation

    E-Print Network [OSTI]

    California at Santa Barbara, University of

    Computing Facilities Orientation September, 2014 #12;Introductions Jason Simpson ­ Manager Computing Facilities Use Policy The Computing facilities are a shared resource for all Bren MESM students Respect the work environment of other students Protect the computer equipment and resources provided You

  19. DOE Facility Management Contracts Facility Owner Contractor

    Broader source: Energy.gov (indexed) [DOE]

    periods 122015 Facility Mgmt 2009 http:www.hanford.govpage.cfmDOEORPContracts Marc McCusker 509-376-2760 Susan E. Bechtol 509-376-3388 Strategic Petroleum Reserve FE Dyn...

  20. UNIVERSITY OF WASHINGTON FINANCE & FACILITIES

    E-Print Network [OSTI]

    Van Volkenburgh, Elizabeth

    UNIVERSITY OF WASHINGTON FINANCE & FACILITIES Capital Projects Office TITLE UNIVERSITY OF WASHINGTON FINANCE & FACILITIES Capital Projects Office UNIVERSITY OF WASHINGTON FINANCE & FACILITIES Capital, 2013 #12;UNIVERSITY OF WASHINGTON FINANCE & FACILITIES Capital Projects Office TITLE · 3.15-mile

  1. Advanced Materials Facilities & Capabilites | ORNL

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Sciences Building Battery Processing Facility Battery and Capacitor Test Facility Nuclear Analytical Chemistry and Isotopics Laboratories Manufacturing Manufacturing Demonstration...

  2. Liquid-Liquid Extraction Equipment

    SciTech Connect (OSTI)

    Jack D. Law; Terry A. Todd

    2008-12-01

    Solvent extraction processing has demonstrated the ability to achieve high decontamination factors for uranium and plutonium while operating at high throughputs. Historical application of solvent extraction contacting equipment implies that for the HA cycle (primary separation of uranium and plutonium from fission products) the equipment of choice is pulse columns. This is likely due to relatively short residence times (as compared to mixer-settlers) and the ability of the columns to tolerate solids in the feed. Savannah River successfully operated the F-Canyon with centrifugal contactors in the HA cycle (which have shorter residence times than columns). All three contactors have been successfully deployed in uranium and plutonium purification cycles. Over the past 20 years, there has been significant development of centrifugal contactor designs and they have become very common for research and development applications. New reprocessing plants are being planned in Russia and China and the United States has done preliminary design studies on future reprocessing plants. The choice of contactors for all of these facilities is yet to be determined.

  3. Characterization, minimization and disposal of radioactive, hazardous, and mixed wastes during cleanup and rransition of the Tritium Research Laboratory (TRL) at Sandia National Laboratories/California (SNL/CA)

    SciTech Connect (OSTI)

    Garcia, T.B.; Gorman, T.P.

    1996-12-01

    This document provides an outline of waste handling practices used during the Sandia National Laboratory/California (SNL/CA), Tritium Research Laboratory (TRL) Cleanup and Transition project. Here we provide background information concerning the history of the TRL and the types of operations that generated the waste. Listed are applicable SNL/CA site-wide and TRL local waste handling related procedures. We describe personnel training practices and outline methods of handling and disposal of compactible and non-compactible low level waste, solidified waste water, hazardous wastes and mixed wastes. Waste minimization, reapplication and recycling practices are discussed. Finally, we provide a description of the process followed to remove the highly contaminated decontamination systems. This document is intended as both a historical record and as a reference to other facilities who may be involved in similar work.

  4. Underground Test Area Subproject Phase I Data Analysis Task. Volume VII - Tritium Transport Model Documentation Package

    SciTech Connect (OSTI)

    1996-12-01

    Volume VII of the documentation for the Phase I Data Analysis Task performed in support of the current Regional Flow Model, Transport Model, and Risk Assessment for the Nevada Test Site Underground Test Area Subproject contains the tritium transport model documentation. Because of the size and complexity of the model area, a considerable quantity of data was collected and analyzed in support of the modeling efforts. The data analysis task was consequently broken into eight subtasks, and descriptions of each subtask's activities are contained in one of the eight volumes that comprise the Phase I Data Analysis Documentation.

  5. Measurement of the electron antineutrino mass from the beta spectrum of gaseous tritium

    SciTech Connect (OSTI)

    Knapp, D.A.

    1986-12-01

    A measurement has been made of the mass of the electron antineutrino using the beta spectrum from a source of gaseous molecular tritium, and an upper limit of 36 eV/c/sup 2/ has been set on this mass. This measurement is the first upper limit on neutrino mass that does not rely on assumptions about the atomic configuration after the beta decay, and it has significantly smaller systematic errors associated with it than do previous measurements. 130 refs., 83 figs., 8 tabs.

  6. RF system considerations for accelerator production of tritium and the transmutation of nuclear waste

    SciTech Connect (OSTI)

    Tallerico, P.J.; Lynch, M.T.

    1993-11-01

    RF driven proton accelerators for the transmutation of nuclear waste (ATW) or for the production of tritium (APT) require unprecedented amounts of CW RF power at UHF frequencies. For both systems, the baseline design is for 246 MW at 700 MHz and 8,5 MW at 350 MHz. The main technical challenges are how to design and build such a large system so that it has excellent reliability, high efficiency, and reasonable capital cost. The issues associated with the selection of the RF amplifier and the sizes of the power supplies are emphasized in this paper.

  7. MICROSTRUCTURAL EXAMINATION AND DEUTERIUM PERMEATION TESTING OF ADVANCED COATINGS FOR TRITIUM SERVICE

    SciTech Connect (OSTI)

    Korinko, P.

    2004-01-24

    A plant directed research and development task to develop and study new, improved, and low cost tritium permeation barriers was initiated in FY02. The project was intended to determine the permeation rate and permeation reduction factor of substrate materials and coated materials. The samples were characterized for microstructural and microchemical consistency. Permeation tests were also run. The sample geometry and sample sealing method selected for the coatings posed significant schedule and technical challenges. Diffusivity were consistent with published values but permeation data exhibited an unexpected sample to sample variation. The effort has lead to an improved sample design that will be used to support a Process Development task.

  8. MICROSTRUCTURAL FEATURES AFFECTING PROPERTIES AND AGING OF TRITIUM-EXPOSED AUSTENTIC STAINLESS STEEL

    SciTech Connect (OSTI)

    Subramanian, K; Michael Morgan, M

    2004-01-10

    A project to implement a life-cycle engineering approach to tritium reservoirs has been initiated through the DOE - Technology Investment Projects. The first task in the project was to develop a comprehensive list of microstructural features that impact the aging performance of the tritium reservoirs. Each of the participating sites (SRNL, SNL, LANL, KCP) independently developed a list of features deemed integral to tritium reservoir performance based upon operational and design experience. An integrated list of features was ultimately developed by the project team that could be included in the modeling process. The features of interest were chosen based upon their impact on the following key factors in controlling crack growth: (1) the H/He solubility or diffusivity within the materials, (2) the stress/strain state at the crack tip, (3) material threshold for crack extension, and (4) microstructure based fracture distance, commonly estimated by grain size for intergranular fracture. Wherever possible, key references were identified to substantiate the effects on the tritium embrittlement phenomenon of the various microstructural features. Each of these features was chosen based upon their impact to the cracking phenomenon of interest. The features chosen were typically associated with orientation, morphology, and distribution of phases and inclusions, grain and grain boundary characteristics, and initial mechanical properties. Phase and inclusion content and distribution were determined to play a key role in the cracking phenomenon. The presence of {delta}-ferrite in the weld and strain-induced martensite in the primarily austenitic matrix are known to facilitate hydrogen diffusion and the interfaces have been observed as a hydrogen assisted fracture path. The morphology, size, and distribution of inclusions and precipitates, particularly on the grain boundaries, influence cracking since they trap hydrogen and facilitate intergranular fracture. Compositional banding and nitrogen concentration were also included as features of interest. The microstructural features of interest included (1) grain size, shape, and orientation; (2) dislocation structure and distribution, or recovered vs. un-recovered. The grain size and orientation affect the grain boundary fracture stress and the hydrogen solubility and diffusion paths. The dislocation structure and distribution play a role in hydrogen trapping as well as potentially affecting the hydrogen assisted fracture path. The initial mechanical and physical properties that are to be included in the investigation are yield stress, fracture toughness, work-hardening capacity, threshold hydrogen cracking stress intensity and stacking-fault energy.

  9. EIS-0288-S1: Production of Tritium in a Commercial Light Water Reactor

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:FinancingPetroleum Based|Department of5 PeerRecord of3: SupplementDisposition(CLWR) Tritium

  10. Facility Effluent Monitoring Plan determinations for the 600 Area facilities

    SciTech Connect (OSTI)

    Nickels, J.M.

    1991-08-01

    This document determines the need for Facility Effluent Monitoring Plans for Westinghouse Hanford Company's 600 Area facilities on the Hanford Site. The Facility Effluent Monitoring Plan determinations were prepared in accordance with A Guide For Preparing Hanford Site Facility Effluent Monitoring Plans (WHC 1991). Five major Westinghouse Hanford Company facilities in the 600 Area were evaluated: the Purge Water Storage Facility, 212-N, -P, and -R Facilities, the 616 Facility, and the 213-J K Storage Vaults. Of the five major facilities evaluated in the 600 Area, none will require preparation of a Facility Effluent Monitoring Plan.

  11. Polyacidic multiloading metal extractants 

    E-Print Network [OSTI]

    Gordon, R. J.; Campbell, J.; Henderson, D.K.; Henry, D. C. R.; Swart, R. M.; Tasker, P. A.; White, F. J.; Wood, J. L.; Yellowlees, L. J

    2008-01-01

    Novel polynucleating, di- and tri-acidic ligands have been designed to increase the molar and mass transport efficiencies for the recovery of base metals by solvent extraction.

  12. Fission Product Extraction Process

    ScienceCinema (OSTI)

    None

    2013-05-28

    A new INL technology can simultaneously extract cesium and strontium for reuse. For more information about INL research, visit http://www.facebook.com/idahonationallaboratory.

  13. Poultry Facility Biosecurity 

    E-Print Network [OSTI]

    Carey, John B.; Prochaska, J. Fred; Jeffrey, Joan S.

    2005-12-21

    . When teamed with disinfection and sanitation pro - cedures, biosecurity practices can eradicate or reduce pathogens to noninfectious levels. Such preventive measures as vaccination and sero- logic monitoring also help ensure good f_lock health... economically, reducing production over the life of the facility without overt signs of disease. Once contaminated with pathogens, poultry facilities are extremely diff_icult and expensive to clean, sanitize and disinfect. Facility location and design...

  14. ARM Mobile Facilities

    SciTech Connect (OSTI)

    Orr, Brad; Coulter, Rich

    2010-12-13

    This video provides an overview of the ARM Mobile Facilities, two portable climate laboratories that can deploy anywhere in the world for campaigns of at least six months.

  15. ARM Mobile Facilities

    ScienceCinema (OSTI)

    Orr, Brad; Coulter, Rich

    2014-09-15

    This video provides an overview of the ARM Mobile Facilities, two portable climate laboratories that can deploy anywhere in the world for campaigns of at least six months.

  16. Presented by FACILITIES MANAGEMENT

    E-Print Network [OSTI]

    Meyers, Steven D.

    Presented by FACILITIES MANAGEMENT TRANSFORMING USF'S TAMPA CAMPUS SUMMER 2011 #12; WELCOME Facili:es Management #12; Facili:es Management #12; NEW CONSTRUCTION Facili

  17. Facility Survey & Transfer

    Broader source: Energy.gov [DOE]

    As DOE facilities become excess, many that are radioactively and/or chemically contaminated will become candidate for transfer to DOE-EM for deactivation and decommissioning.

  18. Industrial application of GNEP solvent-extraction processes

    SciTech Connect (OSTI)

    Arm, S.T.; Phillips, C.; Dobson, A.

    2008-07-01

    EnergySolutions is currently studying the feasibility of commercially recycling spent nuclear fuel in the USA as part of the Global Nuclear Energy Partnership. Uranium, plutonium, and neptunium recycling are accomplished by employing well-established solvent-extraction technology based on the tributylphosphate extractant and acetohydroxamic complexant stripping in a commercially demonstrated configuration. Americium and curium recycling is best achieved by employing the TRUEX and TALSPEAK solvent-extraction processes or a simplified variant of them. Facility design is not predicated on performing any research and development a priori. Process development and demonstration will proceed in parallel with design by proven design-management techniques. (authors)

  19. Texas Facilities Commission's Facility Management Strategic Plan 

    E-Print Network [OSTI]

    Ramirez, J. A.

    2009-01-01

    stream_source_info ESL-IC-09-11-12.pdf.txt stream_content_type text/plain stream_size 4735 Content-Encoding ISO-8859-1 stream_name ESL-IC-09-11-12.pdf.txt Content-Type text/plain; charset=ISO-8859-1 Texas Facilities... Commission?s Facility Management Strategic Plan Jorge A. Ramirez Deputy Executive Director Building Operations & Plant Management ESL-IC-09-11-12 Proceedings of the Ninth International Conference for Enhanced Building Operations, Austin, Texas, November 17...

  20. Recent palladium membrane reactor development at the tritium systems test assembly

    SciTech Connect (OSTI)

    Willms, R.S.; Birdsell, S.A.; Wilhelm, R.C.

    1995-07-01

    The palladium membrane reactor (PMR) is proving to be a simple and effective means for recovering hydrogen isotopes from fusion fuel impurities such as methane and water. This device directly combines two techniques which have long been utilized for hydrogen processing, namely catalytic shift reactions and palladium/silver permeators. A proof-of-principle (PMR) has been constructed and tested at the Tritium Systems Test Assembly of Los Alamos National Laboratory. The first tests with this device showed that is was effective for the proposed purpose. Initial work concluded that a nickel catalyst was an appropriate choice for use in a PMR. More detailed testing of the PMR with such a catalyst was performed and reported in other works. It was shown that a nickel catalyst-packed PMR did, indeed, recover hydrogen from water and methane with efficiencies approaching 100% in a single processing pass. These experiments were conducted over an extended period of time and no failure or need for regeneration was encountered. These positive results have prompted further PMR development. Topics addressed include alternate PMR geometries and initial testing of the PMR with tritium. These are the subjects of this paper.

  1. MINERAL FACILITIES MAPPING PROJECT

    E-Print Network [OSTI]

    Gilbes, Fernando

    Questionnaires. Update the data that pertaining to MIT's contacts worldwide. #12;BOJNOURD CEMENT PLANT Location a database using the Structural Table of Mineral Industry, which includes the location of main mineral The mineral facilities database included: Type of facility: Mine (open pit, underground) Plant ( refineries

  2. Geophysical InversionFacility

    E-Print Network [OSTI]

    Oldenburg, Douglas W.

    UBC Geophysical InversionFacility Modelling and Inversion of EMI data collected over magnetic soils of EMI data acquired at sites with magnetic soils · Geophysical Proveouts · Geonics EM63 Data · First model parameters: · Location · Orientation · Polarizabilities 4 #12;UBC Geophysical Inversion Facility

  3. Nanotechnology User Facility for

    E-Print Network [OSTI]

    A National Nanotechnology User Facility for Industry Academia Government #12;In the NanoFab, you measurement and fabrication methods in response to national nanotechnology needs. www.nist.gov/cnst Robert) is the Department of Commerce's nanotechnology user facility. The CNST enables innovation by providing rapid access

  4. Emergency Facilities and Equipment

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1997-08-21

    This volume clarifies requirements of DOE O 151.1 to ensure that emergency facilities and equipment are considered as part of emergency management program and that activities conducted at these emergency facilities are fully integrated. Canceled by DOE G 151.1-4.

  5. Science &Technology Facilities Council

    E-Print Network [OSTI]

    Science &Technology Facilities Council Accelerator Science and Technology Centre Daresbury Science)1235 445808 www.stfc.ac.uk/astec Head office, Science and Technology Facilities Council, Polaris House, North Newton Group, La Palma: Joint Astronomy Centre, Hawaii. ASTeC Science Highlights 2009 - 2010 Science

  6. Kauai Test Facility

    SciTech Connect (OSTI)

    Hay, R.G.

    1982-01-01

    The Kauai Test Facility (KTF) is a Department of Energy rocket launch facility operated by Sandia National Laboratories. Originally it was constructed in support of the high altitude atmospheric nuclear test phase of operation Dominic in the early 1960's. Later, the facility went through extensive improvement and modernization to become an integral part of the Safeguard C readiness to resume nuclear testing program. Since its inception and build up, in the decade of the sixties and the subsequent upgrades of the seventies, range test activities have shifted from full scale test to emphasis on research and development of materials and components, and to making high altitude scientific measurements. Primarily, the facility is intended to be utilized in support of development programs at the DOE weapons laboratories, however, other organizations may make use of the facility on a non-interface basis. The physical components at KTF and their operation are described.

  7. Facility deactivation and demolition

    SciTech Connect (OSTI)

    Cormier, S.L.; Adamowski, S.J.

    1994-12-31

    Today an improperly closed facility can be a liability to its owner, both financially and environmentally. A facility deactivation program must be planned and implemented to decrease liabilities, minimize operating costs, seek to reuse or sell processes or equipment, and ultimately aid in the sale and/or reuse of the facility and property whether or not the building(s) are demolished. These programs should be characterized within the deactivation plan incorporating the following major categories: Utility Usage; Environmental Decontamination; Ongoing Facility Management; Property Management/Real Estate Issues. This paper will outline the many facets of the facility deactivation and demolition programs implemented across the country for clients in the chemical, automotive, transportation, electronic, pharmaceutical, power, natural gas and petroleum industries. Specific emphasis will be placed on sampling and analysis plans, specification preparation, equipment and technologies utilized, ``how clean is clean`` discussions and regulatory guidelines applicable to these issues.

  8. Data Summary Report for the 1997 Semiannual Tritium Survey for Fourmile Branch and the F- and H-Area Seeplines

    SciTech Connect (OSTI)

    Koch, J.W. II

    1998-01-05

    This report presents a summary of the definitive data validation and verification for the 1997 RFI/RI semiannual tritium survey for Fourmile Branch and the F- and H-Area Seeplines. The RFI/RI was performed under the direction of WSRC ESS/Ecology. This report was prepared under the direction EPD/EMS.

  9. Data Summary Report for the 1998 Semiannual Tritium Surveys for Fourmile Branch and the F- and H-Area Seeplines

    SciTech Connect (OSTI)

    Koch, J.

    1999-01-26

    This report presents a summary of the definitive data validation and verification for the 1998 semiannual tritium surveys for Fourmile Branch and the F- and H-Area Seeplines. The project was performed under the direction of WSRC EAS/Ecology. This report was prepared under the direction of EPD/EMS.

  10. The Tritium Breeding Reality and Need for Near-Term Breeding-Related R&D Programs

    E-Print Network [OSTI]

    tritium breeding ratio (TBR) in the presence of several design elements that compete for the best of how the individual design elements degrade the TBR and what conditions or changes are more damaging the gap, improve the prediction of the minimum required TBR, and develop design elements that help

  11. Bl k t T h l F l C l dBlanket Technology, Fuel Cycle and Tritium Self Sufficiency

    E-Print Network [OSTI]

    California at Los Angeles, University of

    Bl k t T h l F l C l dBlanket Technology, Fuel Cycle and Tritium Self Sufficiency M h d Abd and Technology Center (UCLA) President Council of Energy Research and Education Leaders CEREL (USA functions and multiple materials in multiple-field environment. ­ Multiple effects and synergistic phenomena

  12. Relativistic cyclotron radiation detection of tritium decay electrons as a new technique for measuring the neutrino mass

    E-Print Network [OSTI]

    Monreal, Benjamin

    The shape of the beta-decay energy distribution is sensitive to the mass of the electron neutrino. Attempts to measure the endpoint shape of tritium decay have so far seen no distortion from the zero-mass form, thus placing ...

  13. Ultra-high tritium decontamination of simulated fusion fuel exhaust using a 2-stage palladium membrane reactor

    SciTech Connect (OSTI)

    Birdsell, S.A.; Willms, R.S.; Wilhelm, R.C.

    1996-12-31

    A 2-stage cold (non-tritium) PMR system was tested with the ITER mix in61 days of continuous operation. No decrease in performance was observed over the duration of the test. Decontamination factor (DF) was found to increase with decreasing inlet rate. Decontamination factors in excess of 1.4 {times} 10{sup 5} were obtained, but the exact value of the highest DF could not be determined because of analysis limitations. Results of the 61-day test were used to design a 2-stage PMR system for use in tritium testing. The PMR system was scaled up by a factor of 6 and built into a glovebox in the Tritium Systems Test Assembly (TSTA) of the Los Alamos National Laboratory. This system is approximately 1/5th of the expected full ITER scale. The ITER mix was injected into the PMR system for 31 hours, during which 4.5 g of tritium were processed. The 1st stage had DF = 200 and the 2nd stage had DF = 2.9 {times} 10{sup 6}. The overall DF = 5.8 {times} 10{sup 8}, which is greater than ITER requirements.

  14. THE EFFECTS OF HYDROGEN, TRITIUM, AND HEAT TREATMENT ON THE DEFORMATION AND FRACTURE TOUGHNESS PROPERTIES OF STAINLESS STEEL

    SciTech Connect (OSTI)

    Morgan, M.; Tosten, M.; Chapman, G.

    2013-09-06

    The deformation and fracture toughness properties of forged stainless steels pre-charged with tritium were compared to the deformation and fracture toughness properties of the same steels heat treated at 773 K or 873 K and precharged with hydrogen. Forged stainless steels pre-charged with tritium exhibit an aging effect: Fracture toughness values decrease with aging time after precharging because of the increase in concentration of helium from tritium decay. This study shows that forged stainless steels given a prior heat treatment and then pre-charged with hydrogen also exhibit an aging effect: Fracture toughness values decrease with increasing time at temperature. A microstructural analysis showed that the fracture toughness reduction in the heat-treated steels was due to patches of recrystallized grains that form within the forged matrix during the heat treatment. The combination of hydrogen and the patches of recrystallized grains resulted in more deformation twinning. Heavy deformation twinning on multiple slip planes was typical for the hydrogen-charged samples; whereas, in the non-charged samples, less twinning was observed and was generally limited to one slip plane. Similar effects occur in tritium pre-charged steels, but the deformation twinning is brought on by the hardening associated with decay helium bubbles in the microstructure.

  15. Tritium behavior in eroded dust and debris of plasma-facing A. Hassanein a,*, B. Wiechers b

    E-Print Network [OSTI]

    Harilal, S. S.

    also be redeposited in dierent chemical forms (for example, CT4). There- fore, new materials can properties of these new materials are quite dierent from those of initial PFM. One must also realizeTritium behavior in eroded dust and debris of plasma-facing materials A. Hassanein a,*, B. Wiechers

  16. PPPL3253, Preprint: June 1997, UC420, 426, 427 Alpha Particle Loss in TFTR DeuteriumTritium Plasmas

    E-Print Network [OSTI]

    reduced energy and particle transport of the thermal ions. However, these same conditions reduce the efficiency of plasma heating by the alpha particles and other heating schemes involving fast­tritium (DT) simulations of TFTR deuterium­only experiments. They are compared to results of measurements made

  17. PPPL-3253, Preprint: June 1997, UC-420, 426, 427 Alpha Particle Loss in TFTR Deuterium-Tritium Plasmas

    E-Print Network [OSTI]

    reduced energy and particle transport of the thermal ions. However, these same conditions reduce the efficiency of plasma heating by the alpha particles and other heating schemes involving fast-tritium (DT) simulations of TFTR deuterium-only experiments. They are compared to results of measurements made

  18. Diagnosing fuel R and R asymmetries in cryogenic deuterium-tritium implosions using charged-particle spectrometry at OMEGA

    E-Print Network [OSTI]

    Diagnosing fuel R and R asymmetries in cryogenic deuterium-tritium implosions using charged; published online 22 April 2009 Determining fuel areal density R in moderate- R 100­200 mg/cm2 cryogenic-on deuterons KO-Ds , elastically scattered by primary DT neutrons, from which a fuel R can be inferred

  19. Tritium 2016

    Office of Environmental Management (EM)

    Germany Fus. Sci. Tech 48, 1 (2005) * 8 th 2007 Rochester, USA Fus. Sci. Tech 54, 1&2 (2008) * 9 th 2010 Nara, Japan Fus. Sci. Tech 60, 3&4 (2010) * 10 th 2013 Nice,...

  20. TRITIUM 2013

    Office of Environmental Management (EM)

    www-fusion-magnetique.cea.frtritium2013index.html 1. Containment, safety, and environmental impact 2. Decontamination and waste management 3. Water and air detritiation 4....

  1. tritium1120

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustmentsShirley Ann Jackson About1996HowFOAShowingFuelWeatherizeeEnergyMonumentWestSUMMARY REPORT ON

  2. Short-baseline electron neutrino disappearance, tritium beta decay, and neutrinoless double-beta decay

    SciTech Connect (OSTI)

    Giunti, Carlo; Laveder, Marco [INFN, Sezione di Torino, Via P. Giuria 1, I-10125 Torino (Italy); Dipartimento di Fisica G. Galilei, Universita di Padova, and INFN, Sezione di Padova, Via F. Marzolo 8, I-35131 Padova (Italy)

    2010-09-01

    We consider the interpretation of the MiniBooNE low-energy anomaly and the gallium radioactive source experiments anomaly in terms of short-baseline electron neutrino disappearance in the framework of 3+1 four-neutrino mixing schemes. The separate fits of MiniBooNE and gallium data are highly compatible, with close best-fit values of the effective oscillation parameters {Delta}m{sup 2} and sin{sup 2}2{theta}. The combined fit gives {Delta}m{sup 2}(greater-or-similar sign)0.1 eV{sup 2} and 0.11(less-or-similar sign)sin{sup 2}2{theta}(less-or-similar sign)0.48 at 2{sigma}. We consider also the data of the Bugey and Chooz reactor antineutrino oscillation experiments and the limits on the effective electron antineutrino mass in {beta} decay obtained in the Mainz and Troitsk tritium experiments. The fit of the data of these experiments limits the value of sin{sup 2}2{theta} below 0.10 at 2{sigma}. Considering the tension between the neutrino MiniBooNE and gallium data and the antineutrino reactor and tritium data as a statistical fluctuation, we perform a combined fit which gives {Delta}m{sup 2}{approx_equal}2 eV and 0.01(less-or-similar sign)sin{sup 2}2{theta}(less-or-similar sign)0.13 at 2{sigma}. Assuming a hierarchy of masses m{sub 1}, m{sub 2}, m{sub 3}<tritium data with different mixings in the neutrino and antineutrino sectors. We find a 2.6{sigma} indication of a mixing angle asymmetry.

  3. Superior Energy Performance Industrial Facility Best Practice...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Industrial Facility Best Practice Scorecard Superior Energy Performance Industrial Facility Best Practice Scorecard Superior Energy Performance logo Industrial facilities seeking...

  4. Honda: North American Manufacturing Facilities | Department of...

    Office of Environmental Management (EM)

    Honda: North American Manufacturing Facilities Honda: North American Manufacturing Facilities From October, 2008 Honda: North American Manufacturing Facilities More Documents &...

  5. Liquid chromatographic extraction medium

    DOE Patents [OSTI]

    Horwitz, E.P.; Dietz, M.L.

    1994-09-13

    A method and apparatus are disclosed for extracting strontium and technetium values from biological, industrial and environmental sample solutions using a chromatographic column. An extractant medium for the column is prepared by generating a solution of a diluent containing a Crown ether and dispersing the solution on a resin substrate material. The sample solution is highly acidic and is introduced directed to the chromatographic column and strontium or technetium is eluted using deionized water. 1 fig.

  6. Liquid chromatographic extraction medium

    DOE Patents [OSTI]

    Horwitz, E. Philip (Naperville, IL); Dietz, Mark L. (Evanston, IL)

    1994-01-01

    A method and apparatus for extracting strontium and technetium values from biological, industrial and environmental sample solutions using a chromatographic column is described. An extractant medium for the column is prepared by generating a solution of a diluent containing a Crown ether and dispersing the solution on a resin substrate material. The sample solution is highly acidic and is introduced directed to the chromatographic column and strontium or technetium is eluted using deionized water.

  7. Sandia Energy - About the Facility

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Facility About the FacilityTara Camacho-Lopez2015-05-11T19:38:37+00:00 Test-Bed Wind Turbines Allow Facility Flexibility While Providing Reliable Data in Many Regimes SWiFT will...

  8. Hanford Site Near-Facility Environmental Monitoring Data Report for Calendar Year 2008

    SciTech Connect (OSTI)

    Perkins, Craig J.; Dorsey, Michael C.; Mckinney, Stephen M.; Wilde, Justin W.; Poston, Ted M.

    2009-09-15

    Near-facility environmental monitoring is defined as monitoring near facilities that have the potential to discharge or have discharged, stored, or disposed of radioactive or hazardous materials. Monitoring locations are associated with nuclear facilities such as the Plutonium Finishing Plant, Canister Storage Building, and the K Basins; inactive nuclear facilities such as N Reactor and the Plutonium-Uranium Extraction (PUREX) Facility; and waste storage or disposal facilities such as burial grounds, cribs, ditches, ponds, tank farms, and trenches. Much of the monitoring consists of collecting and analyzing environmental samples and methodically surveying areas near facilities. The program is also designed to evaluate acquired analytical data, determine the effectiveness of facility effluent monitoring and controls, assess the adequacy of containment at waste disposal units, and detect and monitor unusual conditions.

  9. Hanford Site Near-Facility Environmental Monitoring Data Report for Calendar Year 2007- Appendix 2

    SciTech Connect (OSTI)

    Perkins, Craig J.; Dorsey, Michael; Mckinney, Stephen M.; Wilde, Justin W.; Duncan, Joanne P.

    2008-10-13

    Near-facility environmental monitoring is defined as monitoring near facilities that have the potential to discharge or have discharged, stored, or disposed of radioactive or hazardous materials. Monitoring locations are associated with nuclear facilities such as the Plutonium Finishing Plant (PFP), Canister Storage Building (CSB), and the K Basins; inactive nuclear facilities such as N Reactor and the Plutonium-Uranium Extraction (PUREX) Facility; and waste storage or disposal facilities such as burial grounds, cribs, ditches, ponds, tank farms, and trenches. Much of the monitoring consists of collecting and analyzing environmental samples and methodically surveying areas near facilities. The program is also designed to evaluate acquired analytical data, determine the effectiveness of facility effluent monitoring and controls, assess the adequacy of containment at waste disposal units, and detect and monitor unusual conditions.

  10. Kiefer Landfill Biomass Facility | Open Energy Information

    Open Energy Info (EERE)

    Kiefer Landfill Biomass Facility Jump to: navigation, search Name Kiefer Landfill Biomass Facility Facility Kiefer Landfill Sector Biomass Facility Type Landfill Gas Location...

  11. Facility effluent monitoring plan determinations for the 300 Area facilities

    SciTech Connect (OSTI)

    Nickels, J.M.

    1991-08-01

    Facility Effluent Monitoring Plan determinations were conducted for the Westinghouse Hanford Company 300 Area facilities on the Hanford Site. These determinations have been prepared in accordance with A Guide For Preparing Hanford Site Facility Effluent Monitoring Plans. Sixteen Westinghouse Hanford Company facilities in the 300 Area were evaluated: 303 (A, B, C, E, F, G, J and K), 303 M, 306 E, 308, 309, 313, 333, 334 A, and the 340 Waste Handling Facility. The 303, 306, 313, 333, and 334 facilities Facility Effluent Monitoring Plan determinations were prepared by Columbia Energy and Environmental Services of Richland, Washington. The 340 Central Waste Complex determination was prepared by Bovay Northwest, Incorporated. The 308 and 309 facility determinations were prepared by Westinghouse Handford Company. Of the 16 facilities evaluated, 3 will require preparation of a Facility effluent Monitoring Plan: the 313 N Fuels Fabrication Support Building, 333 N Fuels fabrication Building, and the 340 Waste Handling Facility. 26 refs., 5 figs., 10 tabs.

  12. Constraining Mass Spectra with Sterile Neutrinos from Neutrinoless Double Beta Decay, Tritium Beta Decay and Cosmology

    E-Print Network [OSTI]

    Srubabati Goswami; Werner Rodejohann

    2006-05-18

    We analyze the constraints on neutrino mass spectra with extra sterile neutrinos as implied by the LSND experiment. The various mass related observables in neutrinoless double beta decay, tritium beta decay and cosmology are discussed. Both neutrino oscillation results as well as recent cosmological neutrino mass bounds are taken into account. We find that some of the allowed mass patterns are severely restricted by the current constraints, in particular by the cosmological constraints on the total sum of neutrino masses and by the non-maximality of the solar neutrino mixing angle. Furthermore, we estimate the form of the four neutrino mass matrices and also comment on the situation in scenarios with two additional sterile neutrinos.

  13. Constraining mass spectra with sterile neutrinos from neutrinoless double beta decay, tritium beta decay, and cosmology

    SciTech Connect (OSTI)

    Goswami, Srubabati [Harish-Chandra Research Institute, Chhatnag Road, Jhunsi, Allahabad 211 019 (India); Physik-Department, Technische Universitaet Muenchen, James-Franck-Strasse, D-85748 Garching (Germany); Rodejohann, Werner [Physik-Department, Technische Universitaet Muenchen, James-Franck-Strasse, D-85748 Garching (Germany)

    2006-06-01

    We analyze the constraints on neutrino mass spectra with extra sterile neutrinos as implied by the LSND experiment. The various mass related observables in neutrinoless double beta decay, tritium beta decay and cosmology are discussed. Both neutrino oscillation results as well as recent cosmological neutrino mass bounds are taken into account. We find that some of the allowed mass patterns are severely restricted by the current constraints, in particular, by the cosmological constraints on the total sum of neutrino masses and by the nonmaximality of the solar neutrino mixing angle. Furthermore, we estimate the form of the four neutrino mass matrices and also comment on the situation in scenarios with two additional sterile neutrinos.

  14. Theoretical studies on the stopping power of deuterium-tritium mixed with uranium plasmas for ? particles

    SciTech Connect (OSTI)

    Wang, Zhigang; Fu, Zhen-Guo; Zhang, Ping

    2014-10-15

    The stopping power of a compressed and highly ionized deuterium-tritium (DT) and uranium (U) plasma for ? particles at very high temperatures (T?=?5?keV) is examined theoretically with the dimensional continuation method. We show that with increasing density of U, both the magnitude and width of the resonance peak in the stopping power (as a function of the ? particle energy), increases because of the ions, while the penetration distance of the ? particles decreases. A simple relation of decreasing penetration distance as a function of plasma density is observed, which may be useful for inertial confinement fusion experiments. Moreover, by comparing the results with the case of a DT plasma mixed with beryllium, we find that the effect of a higher Z plasma is stronger, with regard to energy loss as well as the penetration distance of ? particles, than that of a lower Z plasma.

  15. Removing tritium and other impurities during industrial recycling of beryllium from a fusion reactor

    SciTech Connect (OSTI)

    Dylst, K.; Seghers, J.; Druyts, F.; Braet, J. [SCK-CEN, Boeretang 200, 2400 Mol (Belgium)

    2008-07-15

    Recycling beryllium used in a fusion reactor might be a good way to overcome problems related to the disposal of neutron irradiated beryllium. The critical issues for the recycling of used first wall beryllium are the presence of tritium and (transuranic) impurities. High temperature annealing seems to be the most promising technique for detritiation. Purification of the de-tritiated beryllium can be achieved by chlorination of the irradiated beryllium and the subsequent reduction of beryllium chloride to highly pure metallic beryllium. After that, the beryllium can be re-fabricated into first wall tiles via powder metallurgy which is already a mature industrial practice. This paper outlines the path to define the experimental needs for beryllium recycling and tackles problems related to the detritiation and the purification via the chlorine route. (authors)

  16. Assessment of database for interaction of tritium with ITER plasma facing materials

    SciTech Connect (OSTI)

    Dolan, T.J.; Anderl, R.A.

    1994-09-01

    The present work surveys recent literature on hydrogen isotope interactions with Be, SS and Inconels, Cu, C, and V, and alloys of Cu and V. The goals are (1) to provide input to the International Thermonuclear Experimental Reactor (ITER) team to help with tritium source term estimates for the Early Safety and Environmental Characterization Study and (2) to provide guidance for planning additional research that will be needed to fill gaps in the present materials database. Properties of diffusivity, solubility, permeability, chemical reactions, Soret effect, recombination coefficient, surface effects, trapping, porosity, layered structures, interfaces, and oxides are considered. Various materials data are tabulated, and a matrix display shows an assessment of the quality of the data available for each main property of each material. Recommendations are made for interim values of diffusivity and solubility to be used, pending further discussion by the ITER community.

  17. Solid Deuterium-Tritium Surface Roughness In A Beryllium Inertial Confinement Fusion Shell

    SciTech Connect (OSTI)

    Kozioziemski, B J; Sater, J D; Moody, J D; Montgomery, D S; Gautier, C

    2006-04-19

    Solid deuterium-tritium (D-T) fuel layers for inertial confinement fusion experiments were formed inside of a 2 mm diameter beryllium shell and were characterized using phase-contrast enhanced x-ray imaging. The solid D-T surface roughness is found to be 0.4 {micro}m for modes 7-128 at 1.5 K below the melting temperature. The layer roughness is found to increase with decreasing temperature, in agreement with previous visible light characterization studies. However, phase-contrast enhanced x-ray imaging provides a more robust surface roughness measurement than visible light methods. The new x-ray imaging results demonstrate clearly that the surface roughness decreases with time for solid D-T layers held at 1.5 K below the melting temperature.

  18. Pollution Control Facilities (South Carolina)

    Broader source: Energy.gov [DOE]

    For the purpose of this legislation, pollution control facilities are defined as any facilities designed for the elimination, mitigation or prevention of air or water pollution, including all...

  19. Listing of Defense Nuclear Facilities

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Plant Mound Facility Fernald Environmental Management Project Site Pantex Plant Rocky Flats Environmental Technology Site, including the Oxnard Facility Savannah River Site Los...

  20. Facility Modernization Report

    SciTech Connect (OSTI)

    Robinson, D; Ackley, R

    2007-05-10

    Modern and technologically up-to-date facilities and systems infrastructure are necessary to accommodate today's research environment. In response, Lawrence Livermore National Laboratory (LLNL) has a continuing commitment to develop and apply effective management models and processes to maintain, modernize, and upgrade its facilities to meet the science and technology mission. The Facility Modernization Pilot Study identifies major subsystems of facilities that are either technically or functionally obsolete, lack adequate capacity and/or capability, or need to be modernized or upgraded to sustain current operations and program mission. This study highlights areas that need improvement, system interdependencies, and how these systems/subsystems operate and function as a total productive unit. Although buildings are 'grandfathered' in and are not required to meet current codes unless there are major upgrades, this study also evaluates compliance with 'current' building, electrical, and other codes. This study also provides an evaluation of the condition and overall general appearance of the structure.

  1. Liquidity facilities and signaling

    E-Print Network [OSTI]

    Arregui, Nicolás

    2010-01-01

    This dissertation studies the role of signaling concerns in discouraging access to liquidity facilities like the IMF contingent credit lines (CCL) and the Discount Window (DW). In Chapter 1, I analyze the introduction of ...

  2. User Facilities | ORNL

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    prior to granting access to a user facility. User Office User Program Manager Laura Morris Edwards 865.574.2966 Email User Office User Office User Program Manager Laura Morris...

  3. Photovoltaic Research Facilities

    Office of Energy Efficiency and Renewable Energy (EERE)

    The U.S. Department of Energy (DOE) funds photovoltaic (PV) research and development (R&D) at its national laboratory facilities located throughout the country. To encourage further innovation,...

  4. Facilities Management Mike Johnson

    E-Print Network [OSTI]

    Capogna, Luca

    , Design & Construction Services Bob Beeler Director, Facility Operations & Maintenance / Environmental Health & Safety Ron Edwards Director, Utility Operations & Maintenance Scott Turley Director, Business & Distribution Utility Plant Operations Water Treatment Zone C Utility Maintenance (HEAT) Power Distribution

  5. NETL - Fuel Reforming Facilities

    ScienceCinema (OSTI)

    None

    2014-06-27

    Research using NETL's Fuel Reforming Facilities explores catalytic issues inherent in fossil-energy related applications, including catalyst synthesis and characterization, reaction kinetics, catalyst activity and selectivity, catalyst deactivation, and stability.

  6. Painter Greenhouse Guidelines Contact: All emails regarding facilities, facilities equipment, supplies at facilities, or watering

    E-Print Network [OSTI]

    Painter Greenhouse Guidelines Contact: All emails regarding facilities, facilities equipment, supplies at facilities, or watering concerns to both the greenhouse manager, Shane Merrell for the Painter Greenhouses must be generated through Shane Merrell. Keep doors locked at all times. Repairs

  7. THE RADIOLOGICAL RESEARCH ACCELERATOR FACILITY The Radiological Research Accelerator Facility

    E-Print Network [OSTI]

    THE RADIOLOGICAL RESEARCH ACCELERATOR FACILITY 71 The Radiological Research Accelerator Facility the irradiated cells. Both the microbeam and the track segment facilities continue to be utilized in various investigations of this phenomenon. The single- particle microbeam facility provides precise control of the number

  8. THE RADIOLOGICAL RESEARCH ACCELERATOR FACILITY The Radiological Research Accelerator Facility

    E-Print Network [OSTI]

    THE RADIOLOGICAL RESEARCH ACCELERATOR FACILITY 1 The Radiological Research Accelerator Facility for Radiological Research (CRR). Using the mi- crobeam facility, 10% of the cells were irradiated through particle beam as well as the first fo- cused microbeam in the new microbeam facility. · Another significant

  9. Facility Location with Hierarchical Facility Costs Zoya Svitkina #

    E-Print Network [OSTI]

    Tardos, Ã?va

    Facility Location with Hierarchical Facility Costs Zoya Svitkina # â?? Eva Tardos + Abstract We consider the facility location problem with hierarchi­ cal facility costs, and give a (4 installation costs. Shmoys, Swamy and Levi [13] gave an approxi­ mation algorithm for a two­level version

  10. Working with SRNL - Our Facilities - Glovebox Facilities

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantityBonneville Power AdministrationRobust,Field-effectWorking WithTelecentricNCubictheThepresented in1: ModelGlovebox Facilities

  11. Modified Purex first-cycle extraction for neptunium recovery

    SciTech Connect (OSTI)

    Dinh, Binh; Moisy, Philippe; Baron, Pascal; Calor, Jean-Noel; Espinoux, Denis; Lorrain, Brigitte; Benchikouhne-Ranchoux, Magali

    2008-07-01

    A new PUREX first-cycle flowsheet was devised to enhance the extraction yield of neptunium at the extraction step of this cycle. Simulation results (using a qualified process-simulation tool), le d to raising the nitric acid concentration of the feed from 3 M to 4.5 M to allow extraction of more than 99% of the neptunium. This flowsheet was operated in the shielded process cell of ATALANTE facility using pulsed columns and mixer-settlers banks. A 15 kg quantity of genuine oxide fuel of average burn up of 52 GWd/t with cooling time of nearly five years was treated, and the neptunium extraction yield obtained was greater than 99.6%. (authors)

  12. Safety evaluation report related to the Department of Energy`s proposal for the irradiation of lead test assemblies containing tritium-producing burnable absorber rods in commercial light-water reactors. Project Number 697

    SciTech Connect (OSTI)

    1997-05-01

    The NRC staff has reviewed a report, submitted by DOE to determine whether the use of a commercial light-water reactor (CLWR) to irradiate a limited number of tritium-producing burnable absorber rods (TPBARs) in lead test assemblies (LTAs) raises generic issues involving an unreviewed safety question. The staff has prepared this safety evaluation to address the acceptability of these LTAs in accordance with the provision of 10 CFR 50.59 without NRC licensing action. As summarized in Section 10 of this safety evaluation, the staff has identified issues that require NRC review. The staff has also identified a number of areas in which an individual licensee undertaking irradiation of TPBAR LTAs will have to supplement the information in the DOE report before the staff can determine whether the proposed irradiation is acceptable at a particular facility. The staff concludes that a licensee undertaking irradiation of TPBAR LTAs in a CLWR will have to submit an application for amendment to its facility operating license before inserting the LTAs into the reactor.

  13. Extracting the Eliashberg Function

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantityBonneville Power Administration would like submitKansasCommunitiesof Energy8)highlightsNewExtracellularExtractingExtracting

  14. PUREX/UO{sub 3} facilities deactivation lessons learned history

    SciTech Connect (OSTI)

    Hamrick, D.G.; Gerber, M.S.

    1995-01-01

    The Plutonium-Uranium Extraction (PUREX) Facility operated from 1956-1972, from 1983-1988, and briefly during 1989-1990 to produce for national defense at the Hanford Site in Washington State. The Uranium Trioxide (UO{sub 3}) Facility operated at the Hanford Site from 1952-1972, 1984-1988, and briefly in 1993. Both plants were ordered to permanent shutdown by the U.S. Department of Energy (DOE) in December 1992, thus initiating their deactivation phase. Deactivation is that portion of a facility`s life cycle that occurs between operations and final decontamination and decommissioning (D&D). This document details the history of events, and the lessons learned, from the time of the PUREX Stabilization Campaign in 1989-1990, through the end of the first full fiscal year (FY) of the deactivation project (September 30, 1994).

  15. Comprehensive facilities plan

    SciTech Connect (OSTI)

    1997-09-01

    The Ernest Orlando Lawrence Berkeley National Laboratory`s Comprehensive Facilities Plan (CFP) document provides analysis and policy guidance for the effective use and orderly future development of land and capital assets at the Berkeley Lab site. The CFP directly supports Berkeley Lab`s role as a multiprogram national laboratory operated by the University of California (UC) for the Department of Energy (DOE). The CFP is revised annually on Berkeley Lab`s Facilities Planning Website. Major revisions are consistent with DOE policy and review guidance. Facilities planing is motivated by the need to develop facilities for DOE programmatic needs; to maintain, replace and rehabilitate existing obsolete facilities; to identify sites for anticipated programmatic growth; and to establish a planning framework in recognition of site amenities and the surrounding community. The CFP presents a concise expression of the policy for the future physical development of the Laboratory, based upon anticipated operational needs of research programs and the environmental setting. It is a product of the ongoing planning processes and is a dynamic information source.

  16. A Study on a Tritium Separation Process Using Self-Developing Gas Chromatography with Pd-Pt Alloy

    SciTech Connect (OSTI)

    Kojima, S. [JGC Corporation (Japan); Yokosawa, M. [JGC Corporation (Japan); Matsuyama, M. [Toyama University (Japan); Numata, M. [JGC Corporation (Japan); Kato, T. [JGC Corporation (Japan); Watanabe, K. [Toyama University (Japan)

    2005-07-15

    To study the practical application of a tritium separation process using Self-Developing Gas Chromatography (SDGC) using a Pd-Pt alloy, intermediate scale-up experiments (22 mm ID x 2 m length column) and the development of a computational simulation method have been conducted. In addition, intermediate scale production of Pd-Pt powder has been developed for the scale-up experiments.The following results were obtained: (1) a 50-fold scale-up from 3 mm to 22 mm causes no significant impact on the SDGC process; (2) the Pd-Pt alloy powder is applicable to a large size SDGC process; and (3) the simulation enables preparation of a conceptual design of a SDGC process for tritium separation.

  17. Status of the KATRIN experiment and prospects to search for keV-mass sterile neutrinos in tritium ?-decay

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Mertens, Susanne

    2015-03-24

    In this contribution the current status and future perspectives of the Karlsruhe Tritium Neutrino (KATRIN) Experiment are presented. The prime goal of this single ?-decay experiment is to probe the absolute neutrino mass scale with a sensitivity of 200 meV (90% CL). We discuss first results of the recent main spectrometer commissioning measurements, successfully verifying the spectrometer’s basic vacuum, transmission and background properties. We also discuss the prospects of making use of the KATRIN tritium source, to search for sterile neutrinos in the multi-keV mass range constituting a classical candidate for Warm Dark Matter. Due to the very high sourcemore »luminosity, a statistical sensitivity down to active-sterile mixing angles of sin² ? « less

  18. PRODUCTION OF CARBON PRODUCTS USING A COAL EXTRACTION PROCESS

    SciTech Connect (OSTI)

    Dady Dadyburjor; Chong Chen; Elliot B. Kennel; Liviu Magean; Peter G. Stansberry; Alfred H. Stiller; John W. Zondlo

    2006-02-23

    The purpose of this DOE-funded effort is to develop technologies for carbon products from coal-derived feedstocks. Carbon products can include precursor materials such as solvent extracted carbon ore (SECO) and synthetic pitch (Synpitch). In addition, derived products include carbon composites, fibers, foams and others. Key milestones included producing hydrogenated coal in the Hydrotreating Facility for the first time. The facility is now operational, although digital controls have not yet been completely wired. In addition, ultrasound is being used to investigate enhanced dissolution of coal. Experiments have been carried out.

  19. RCRA facility stabilization initiative

    SciTech Connect (OSTI)

    Not Available

    1995-02-01

    The RCRA Facility Stabilization Initiative was developed as a means of implementing the Corrective Action Program`s management goals recommended by the RIS for stabilizing actual or imminent releases from solid waste management units that threaten human health and the environment. The overall goal of stabilization is to, as situations warrant, control or abate threats to human health and/or the environment from releases at RCRA facilities, and/or to prevent or minimize the further spread of contamination while long-term remedies are pursued. The Stabilization initiative is a management philosophy and should not be confused with stabilization technologies.

  20. Facilities | Argonne National Laboratory

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity ofkandz-cm11 Outreach Home Room NewsInformation Current HABFES OctoberEvan Racah861 ANNUAL ELECTRICRashiFacilitiesFacilities

  1. Facilities | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE: Alternative Fuelsof Energy Services » Program ManagementAct4 DOE/CF-0074Facilities Facilities

  2. ALARA engineering at Department of Energy facilities: Bibliography of selected readings in radiation protection and ALARA

    SciTech Connect (OSTI)

    Dionne, B.J.; Khan, T.A.; Lane, S.G.; Baum, J.W.

    1991-05-01

    Promoting the exchange of information related to implementation of the As Low As Reasonably Achievable (ALARA) philosophy is a continuing objective for the Department of Energy (DOE). This report, prepared by the Brookhaven National Laboratory (BNL) ALARA Center for the DOE Office of Health, is the second in a series of bibliographies on dose reduction at DOE facilities. This bibliography contains abstracts relating to various aspects of ALARA program implementation and dose reduction activities, with a specific focus towards DOE facilities. Facility types and activities covered in the scope of this report include: radioactive waste; uranium enrichment; fuel fabrication, storage, and reprocessing; facility decommissioning; hot laboratories; tritium production; research, test and production reactors; weapons fabrication and testing; and accelerators. Material on improved shielding design, decontamination, containments, robotics, job planning, improved operational techniques, and other topics has also been included. This volume (Volume 2 of the series) contains 127 abstracts numbered from 69 through 195, as well as author and subject indices. The subject index contains the abstract numbers from both the previous volume and the current volume, the latter being indicated in boldface. Information that the reader feels should be included in the next volume of this bibliography should be submitted to the BNL ALARA Center.

  3. Comment on "Arsenic Mobility and Groundwater Extraction in

    E-Print Network [OSTI]

    Basu, Asish R.

    -season vertical hydraulic gradients in Bangladesh. Fig. 2 shows weekly measure- ments of water level between 1967 and pre-bomb atmospheric tritium levels of 5 tritium units (TU)], similar to those estimated by Harvey et irrigation pumping took place, clearly indicate that vertical hydraulic gradients and dry-season recharge did

  4. Manufacturing Demonstration Facility

    E-Print Network [OSTI]

    life-cycle energy and greenhouse gas emissions, lower production cost, and create new products Demonstration Facility (865) 574-4351 blueca@ornl.gov INNOVATIONS IN MANUFACTURING www to reduce risk and accelerate the development and deployment of innovative energy-efficient manufacturing

  5. NISCO Cogeneration Facility 

    E-Print Network [OSTI]

    Zierold, D. M.

    1994-01-01

    The NISCO Cogeneration facility utilizes two fluidized bed boilers to generate 200 MW of electricity and up to 80,000 LBS/HR of steam for process use. The partnership, of three industrial electricity users, Citgo, Conoco, and Vista Chemical...

  6. PUREX/UO{sub 3} facilities deactivation lessons learned: History

    SciTech Connect (OSTI)

    Gerber, M.S.

    1997-11-25

    In May 1997, a historic deactivation project at the PUREX (Plutonium URanium EXtraction) facility at the Hanford Site in south-central Washington State concluded its activities (Figure ES-1). The project work was finished at $78 million under its original budget of $222.5 million, and 16 months ahead of schedule. Closely watched throughout the US Department of Energy (DOE) complex and by the US Department of Defense for the value of its lessons learned, the PUREX Deactivation Project has become the national model for the safe transition of contaminated facilities to shut down status.

  7. Nano Research Facility Lab Safety Manual Nano Research Facility

    E-Print Network [OSTI]

    Subramanian, Venkat

    1 Nano Research Facility Lab Safety Manual Nano Research Facility: Weining Wang Office: Brauer---chemical, biological, or radiological. Notify the lab manager, Dr. Yujie Xiong at 5-4530. Eye Contact: Promptly flush

  8. Solid phase extraction membrane

    DOE Patents [OSTI]

    Carlson, Kurt C [Nashville, TN; Langer, Roger L [Hudson, WI

    2002-11-05

    A wet-laid, porous solid phase extraction sheet material that contains both active particles and binder and that possesses excellent wet strength is described. The binder is present in a relatively small amount while the particles are present in a relatively large amount. The sheet material is sufficiently strong and flexible so as to be pleatable so that, for example, it can be used in a cartridge device.

  9. A diamond based neutron spectrometer for diagnostics of deuterium-tritium fusion plasmas

    SciTech Connect (OSTI)

    Cazzaniga, C., E-mail: carlo.cazzaniga@mib.infn.it; Nocente, M.; Gorini, G. [University of Milano Bicocca, Piazza della Scienza 3, Milano (Italy); Istituto di Fisica del Plasma, Associazione EURATOM-ENEA-CNR, via Roberto Cozzi 53, Milano (Italy); Rebai, M.; Giacomelli, L. [University of Milano Bicocca, Piazza della Scienza 3, Milano (Italy); Tardocchi, M.; Croci, G.; Grosso, G. [Istituto di Fisica del Plasma, Associazione EURATOM-ENEA-CNR, via Roberto Cozzi 53, Milano (Italy); Calvani, P.; Girolami, M.; Trucchi, D. M. [CNR-ISM, Research Area Roma 1, Via Salaria km 29.300, 00015-Monterotondo Scalo (Rm) (Italy); Griesmayer, E. [Atominstitut, Vienna University of Technology, Vienna (Austria); Pillon, M. [Associazione EURATOM-ENEA sulla Fusione ENEA C.R. Frascati, Via E. Fermi, 45, 00044 Frascati (Roma) (Italy)

    2014-11-15

    Single crystal Diamond Detectors (SDD) are being increasingly exploited for neutron diagnostics in high power fusion devices, given their significant radiation hardness and high energy resolution capabilities. The geometrical efficiency of SDDs is limited by the size of commercially available crystals, which is often smaller than the dimension of neutron beams along collimated lines of sight in tokamak devices. In this work, we present the design and fabrication of a 14 MeV neutron spectrometer consisting of 12 diamond pixels arranged in a matrix, so to achieve an improved geometrical efficiency. Each pixel is equipped with an independent high voltage supply and read-out electronics optimized to combine high energy resolution and fast signals (<30 ns), which are essential to enable high counting rate (>1 MHz) spectroscopy. The response function of a prototype SDD to 14 MeV neutrons has been measured at the Frascati Neutron Generator by observation of the 8.3 MeV peak from the {sup 12}C(n, ?){sup 9}Be reaction occurring between neutrons and {sup 12}C nuclei in the detector. The measured energy resolution (2.5% FWHM) meets the requirements for neutron spectroscopy applications in deuterium-tritium plasmas.

  10. DYNAMIC ANALYSIS OF THE BULK TRITIUM SHIPPING PACKAGE SUBJECTED TO CLOSURE TORQUES AND SEQUENTIAL IMPACTS

    SciTech Connect (OSTI)

    Wu, T; Paul Blanton, P; Kurt Eberl, K

    2007-07-09

    This paper presents a finite-element technique to simulate the structural responses and to evaluate the cumulative damage of a radioactive material packaging requiring bolt closure-tightening torque and subjected to the scenarios of the Hypothetical Accident Conditions (HAC) defined in the Code of Federal Regulations Title 10 part 71 (10CFR71). Existing finite-element methods for modeling closure stresses from bolt pre-load are not readily adaptable to dynamic analyses. The HAC events are required to occur sequentially per 10CFR71 and thus the evaluation of the cumulative damage is desirable. Generally, each HAC event is analyzed separately and the cumulative damage is partially addressed by superposition. This results in relying on additional physical testing to comply with 10CFR71 requirements for assessment of cumulative damage. The proposed technique utilizes the combination of kinematic constraints, rigid-body motions and structural deformations to overcome some of the difficulties encountered in modeling the effect of cumulative damage. This methodology provides improved numerical solutions in compliance with the 10CFR71 requirements for sequential HAC tests. Analyses were performed for the Bulk Tritium Shipping Package (BTSP) designed by Savannah River National Laboratory to demonstrate the applications of the technique. The methodology proposed simulates the closure bolt torque preload followed by the sequential HAC events, the 30-foot drop and the 30-foot dynamic crush. The analytical results will be compared to the package test data.

  11. Deuterium-Tritium Simulations of the Enhanced Reversed Shear Mode in the Tokamak Fusion Test Reactor

    SciTech Connect (OSTI)

    Mikkelsen, D.R.; Manickam, J.; Scott, S.D.; Zarnstorff

    1997-04-01

    The potential performance, in deuterium-tritium plasmas, of a new enhanced con nement regime with reversed magnetic shear (ERS mode) is assessed. The equilibrium conditions for an ERS mode plasma are estimated by solving the plasma transport equations using the thermal and particle dif- fusivities measured in a short duration ERS mode discharge in the Tokamak Fusion Test Reactor [F. M. Levinton, et al., Phys. Rev. Letters, 75, 4417, (1995)]. The plasma performance depends strongly on Zeff and neutral beam penetration to the core. The steady state projections typically have a central electron density of {approx}2:5x10 20 m{sup -3} and nearly equal central electron and ion temperatures of {approx}10 keV. In time dependent simulations the peak fusion power, {approx} 25 MW, is twice the steady state level. Peak performance occurs during the density rise when the central ion temperature is close to the optimal value of {approx} 15 keV. The simulated pressure profiles can be stable to ideal MHD instabilities with toroidal mode number n = 1, 2, 3, 4 and {infinity} for {beta}{sub norm} up to 2.5; the simulations have {beta}{sub norm} {le} 2.1. The enhanced reversed shear mode may thus provide an opportunity to conduct alpha physics experiments in conditions imilar to those proposed for advanced tokamak reactors.

  12. Accelerator production of tritium 700 MHz and 350 MHz klystron test results

    SciTech Connect (OSTI)

    Rees, D.; Lynch, M.; Tallerico, P.

    1998-12-31

    The Accelerator Production of Tritium project (APT) utilizes a 1,700 MeV, 100 mA proton Linac. The radio frequency (RF) power is provided by 244 continuous wave (CW) klystron amplifiers at 350 MHz and 700 MHz. All but three of the klystrons operate at a frequency of 700 MHz. The 350 MHz klystrons have a nominal output power of 1.2 MW at a DC-to-RF conversion efficiency of 65%. They are modulating-anode klystrons and operate at a beam voltage and current of 95 kV and 20 A. The design is based on the CERN klystron. The 700 MHz klystron is a new development for APT. Three 700 MHz klystrons are currently under development. Two vendors are each developing a baseline klystron that has a nominal output power of 1.0 MW at a DC-to-RF conversion efficiency of 65%. A 700 MHz klystron is also under development that promises to provide an efficiency in excess of 70%. The 700 MHz klystrons operate at a maximum beam voltage of 95 kV and a maximum beam current of 17 A. The test results of these klystrons will be presented and the design features will be discussed.

  13. Viscosity and mutual diffusion of deuterium-tritium mixtures in the warm-dense-matter regime

    SciTech Connect (OSTI)

    Kress, J. D.; Cohen, James S.; Horner, D. A.; Collins, L. A. [Theoretical Division, Los Alamos National Laboratory, Los Alamos, New Mexico 87545 (United States); Lambert, F. [CEA, DAM, DIF, F-91297 Arpajon (France)

    2010-09-15

    We have calculated viscosity and mutual diffusion of deuterium-tritium (DT) in the warm, dense matter regime for densities from 5 to 20 g/cm{sup 3} and temperatures from 2 to 10 eV, using both finite-temperature Kohn-Sham density-functional theory molecular dynamics (QMD) and orbital-free molecular dynamics (OFMD). The OFMD simulations are in generally good agreement with the benchmark QMD results, and we conclude that the simpler OFMD method can be used with confidence in this regime. For low temperatures (3 eV and below), one-component plasma (OCP) model simulations for diffusion agree with the QMD and OFMD calculations, but deviate by 30% at 10 eV. In comparison with the QMD and OFMD results, the OCP viscosities are not as good as for diffusion, especially for 5 g/cm{sup 3} where the temperature dependence is significantly different. The QMD and OFMD reduced diffusion and viscosity coefficients are found to depend largely, though not completely, only on the Coulomb coupling parameter {Gamma}, with a minimum in the reduced viscosity at {Gamma}{approx_equal}25, approximately the same position found in the OCP simulations. The QMD and OFMD equations of state (pressure) are also compared with the hydrogen two-component plasma model.

  14. Elastic properties of Pd-hydrogen, Pd-deuterium, and Pd-tritium single crystals

    SciTech Connect (OSTI)

    Schwarz, R.B. . E-mail: rxzs@lanl.gov; Bach, H.T.; Harms, U.; Tuggle, D.

    2005-02-01

    We used a resonant-ultrasound-spectroscopy technique to measure the three independent elastic constants of PdH{sub x}, PdD{sub x}, and PdT{sub x} single crystals at 300 K. For 0.1x0.62 our PdH{sub x} crystals are two-phase mixtures of coherent {alpha} and {beta} hydride phases. For increasing x in this range, C{sub 44} decreases monotonically whereas C'=12(C11-C12) has a concave parabolic dependence. This difference is because C' is softened by an anelastic relaxation resulting from acoustic-stress-induced changes in the shape of the coherent lenticular-shape precipitates ({beta}-hydride precipitates in {alpha}-hydride matrix and {alpha}-hydride precipitates in {beta}-hydride matrix). In the {beta}-phase C' and C{sub 44} decrease with increasing hydrogen (or deuterium or tritium) content. Furthermore, C' exhibits a strong isotope effect whereas C{sub 44} does not. This effect is attributed to differences in the excitation of optical phonons in Pd-H, Pd-D and Pd-T.

  15. Innovative Drying and Nutrients Extraction

    E-Print Network [OSTI]

    to the extraction process. This method evaporates the water from the products but also drives off up to 70 percent dimethyl ether to extract the water from the material. The new process does not require the addition of heat to evaporate the water during the extraction process. Dimethyl ether has a lower heat

  16. CHEM333: Experiment 2: Extraction

    E-Print Network [OSTI]

    Taber, Douglass

    CHEM­333: Experiment 2: Extraction: Prelab Assignment: Read chapter 4. In this lab you will perform an extraction (Chapter 4; Experiment B). Extraction is one of the easiest purification methods in the organic are insoluble in neutral/acidic water but are soluble in basic water. Follow the protocol and make sure that you

  17. Neutron source reconstruction from pinhole imaging at National Ignition Facility

    SciTech Connect (OSTI)

    Volegov, P.; Danly, C. R.; Grim, G. P.; Guler, N.; Merrill, F. E.; Wilde, C. H.; Wilson, D. C. [Los Alamos National Laboratory, Los Alamos, New Mexico 87544 (United States)] [Los Alamos National Laboratory, Los Alamos, New Mexico 87544 (United States); Fittinghoff, D. N.; Izumi, N.; Ma, T.; Warrick, A. L. [Livermore National Laboratory, Livermore, California 94550 (United States)] [Livermore National Laboratory, Livermore, California 94550 (United States)

    2014-02-15

    The neutron imaging system at the National Ignition Facility (NIF) is an important diagnostic tool for measuring the two-dimensional size and shape of the neutrons produced in the burning deuterium-tritium plasma during the ignition stage of inertial confinement fusion (ICF) implosions at NIF. Since the neutron source is small (?100 ?m) and neutrons are deeply penetrating (>3 cm) in all materials, the apertures used to achieve the desired 10-?m resolution are 20-cm long, single-sided tapers in gold. These apertures, which have triangular cross sections, produce distortions in the image, and the extended nature of the pinhole results in a non-stationary or spatially varying point spread function across the pinhole field of view. In this work, we have used iterative Maximum Likelihood techniques to remove the non-stationary distortions introduced by the aperture to reconstruct the underlying neutron source distributions. We present the detailed algorithms used for these reconstructions, the stopping criteria used and reconstructed sources from data collected at NIF with a discussion of the neutron imaging performance in light of other diagnostics.

  18. Virginia Commonwealth University Facilities Management

    E-Print Network [OSTI]

    Hammack, Richard

    .3 Solid Waste Management 14 018.4 Pest Management Plan 14 Facilities Management Construction & Design Virginia Commonwealth University Facilities Management Construction & Design Construction Management (804) 6285199 VCU Construction & Inspection Management jghosh

  19. Biomass Feedstock National User Facility

    Office of Energy Efficiency and Renewable Energy (EERE)

    Breakout Session 1B—Integration of Supply Chains I: Breaking Down Barriers Biomass Feedstock National User Facility Kevin L. Kenney, Director, Biomass Feedstock National User Facility, Idaho National Laboratory

  20. National Ignition Facility & Photon Science What

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Ignition Facility & Photon Science What is NiF? the national ignition Facility: bringing star Power to earth The National Ignition Facility (NIF) is the world's largest and...

  1. CFTF | Carbon Fiber Technology Facility | ORNL

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    BTRIC CNMS CSMB CFTF Working with CFTF HFIR MDF NTRC OLCF SNS Carbon Fiber Technology Facility Home | User Facilities | CFTF CFTF | Carbon Fiber Technology Facility SHARE Oak...

  2. Proton beam therapy facility

    SciTech Connect (OSTI)

    Not Available

    1984-10-09

    It is proposed to build a regional outpatient medical clinic at the Fermi National Accelerator Laboratory (Fermilab), Batavia, Illinois, to exploit the unique therapeutic characteristics of high energy proton beams. The Fermilab location for a proton therapy facility (PTF) is being chosen for reasons ranging from lower total construction and operating costs and the availability of sophisticated technical support to a location with good access to patients from the Chicago area and from the entire nation. 9 refs., 4 figs., 26 tabs.

  3. THE RADIOLOGICAL RESEARCH ACCELERATOR FACILITY

    E-Print Network [OSTI]

    THE RADIOLOGICAL RESEARCH ACCELERATOR FACILITY #12;115 THE RADIOLOGICAL RESEARCH ACCELERATOR FACILITY An NIH-Supported Resource Center WWW.RARAF.ORG Director: David J. Brenner, Ph.D., D.Sc. Manager delighted that NIH funding for continued development of our single-particle microbeam facility was renewed

  4. Alpha Gamma Hot Cell Facility

    E-Print Network [OSTI]

    Kemner, Ken

    . These operations can result in elevated radiological risks to the facility and workers. ARG-US -- meaning and should be developed for and deployed in nuclear and radiological facilities to aid operation and reduceAlpha Gamma Hot Cell Facility Argonne National Laboratory is a U.S. Department of Energy laboratory

  5. THE RADIOLOGICAL RESEARCH ACCELERATOR FACILITY

    E-Print Network [OSTI]

    175 THE RADIOLOGICAL RESEARCH ACCELERATOR FACILITY #12;176 #12;177 THE RADIOLOGICAL RESEARCH the microbeam and the track-segment facilities have been utilized in various investigations. Table 1 lists-segment facility. Samples are treated with graded doses of radical scavengers to observe changes in the cluster

  6. Facilities Design and Construction Services

    E-Print Network [OSTI]

    Frantz, Kyle J.

    for custodial materials as well as maintenance equipment. (The Facilities Maintenance and Operations Department be in written request to University's Facilities Maintenance and Operations Department and the Communication, corridors and facilities shall provide maximum flexibility and access for routine maintenance. (Reference

  7. Inverse hydrochemical models of aqueous extracts tests

    E-Print Network [OSTI]

    Zheng, L.

    2010-01-01

    years to improve water extraction methods, develop numericalreactions during water extraction, redox processes were notAranyossy, J.F. , 2001. Extraction of water and solutes from

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

    SciTech Connect (OSTI)

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

    1991-11-01

    Promoting the exchange of information related to implementation of the As Low as Reasonably Achievable (ALARA) philosophy is a continuing objective for the Department of Energy (DOE). This report, prepared by the Brookhaven National Laboratory (BNL) ALARA Center for the DOE Office of Health, contains the third in a series of bibliographies on dose reduction at DOE facilities. This report also contains abstracts from the two previous volumes. The BNL ALARA Center was originally established in 1983 under the sponsorship of the Nuclear Regulatory Commission to monitor dose-reduction research and ALARA activities at nuclear power plants. This effort was expanded in 1988 by the DOE's Office of Environment, Safety and Health to include DOE nuclear facilities. This bibliography contains abstracts relating to various aspects of ALARA program implementation and dose-reduction activities, with a specific focus on DOE facilities. Abstracts included in this bibliography were selected from proceedings of technical meetings, journals, research reports, searches of the DOE Energy Data Base, and reprints of published articles provided by the authors. Facility types and activities covered in the scope of this report include: radioactive waste, uranium enrichment, fuel fabrication, storage, and reprocessing, facility decommissioning, hot laboratories, tritium production, research, test and production reactors, weapons fabrication and testing, and accelerators. Material on improved shielding design, decontamination, containments, robotics, job planning, improved operational techniques, and other topics are also included.

  9. Extracting the Eliashberg Function

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantityBonneville Power Administration would like submitKansasCommunitiesof Energy8)highlightsNewExtracellular ProteinsExtracting

  10. Extracting the Eliashberg Function

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantityBonneville Power Administration would like submitKansasCommunitiesof Energy8)highlightsNewExtracellularExtracting the

  11. Extracting the Eliashberg Function

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantityBonneville Power Administration would like submitKansasCommunitiesof Energy8)highlightsNewExtracellularExtracting

  12. Ignition and Inertial Confinement Fusion at The National Ignition Facility

    SciTech Connect (OSTI)

    Moses, E

    2009-10-01

    The National Ignition Facility (NIF), the world's largest and most powerful laser system for inertial confinement fusion (ICF) and for studying high-energy-density (HED) science, is now operational at Lawrence Livermore National Laboratory (LLNL). The NIF is now conducting experiments to commission the laser drive, the hohlraum and the capsule and to develop the infrastructure needed to begin the first ignition experiments in FY 2010. Demonstration of ignition and thermonuclear burn in the laboratory is a major NIF goal. NIF will achieve this by concentrating the energy from the 192 beams into a mm{sup 3}-sized target and igniting a deuterium-tritium mix, liberating more energy than is required to initiate the fusion reaction. NIF's ignition program is a national effort managed via the National Ignition Campaign (NIC). The NIC has two major goals: execution of DT ignition experiments starting in FY2010 with the goal of demonstrating ignition and a reliable, repeatable ignition platform by the conclusion of the NIC at the end of FY2012. The NIC will also develop the infrastructure and the processes required to operate NIF as a national user facility. The achievement of ignition at NIF will demonstrate the scientific feasibility of ICF and focus worldwide attention on laser fusion as a viable energy option. A laser fusion-based energy concept that builds on NIF, known as LIFE (Laser Inertial Fusion Energy), is currently under development. LIFE is inherently safe and can provide a global carbon-free energy generation solution in the 21st century. This paper describes recent progress on NIF, NIC, and the LIFE concept.

  13. UNIVERSITY OF WASHINGTON FINANCE & FACILITIES Faculty Council on University Facilities and Services Presentation October 30, 2014

    E-Print Network [OSTI]

    Van Volkenburgh, Elizabeth

    UNIVERSITY OF WASHINGTON FINANCE & FACILITIES Faculty Council on University Facilities and Services Presentation October 30, 2014 Capital Projects Office TITLE #12;UNIVERSITY OF WASHINGTON FINANCE & FACILITIES Office TITLE #12;UNIVERSITY OF WASHINGTON FINANCE & FACILITIES Faculty Council on University Facilities

  14. Single crystal diamond detector measurements of deuterium-deuterium and deuterium-tritium neutrons in Joint European Torus fusion plasmas

    SciTech Connect (OSTI)

    Cazzaniga, C., E-mail: carlo.cazzaniga@mib.infn.it; Gorini, G.; Nocente, M. [Department of Physics “G. Occhialini,” University of Milano Bicocca, Piazza della Scienza 3, Milano (Italy) [Department of Physics “G. Occhialini,” University of Milano Bicocca, Piazza della Scienza 3, Milano (Italy); Istituto di Fisica del Plasma, Associazione EURATOM-ENEA-CNR, via Roberto Cozzi 53, Milano (Italy); Sundén, E. Andersson; Binda, F.; Ericsson, G. [Department of Physics and Astronomy, EURATOM-VR Association, Uppsala University, Uppsala (Sweden)] [Department of Physics and Astronomy, EURATOM-VR Association, Uppsala University, Uppsala (Sweden); Croci, G.; Grosso, G.; Cippo, E. Perelli; Tardocchi, M. [Istituto di Fisica del Plasma, Associazione EURATOM-ENEA-CNR, via Roberto Cozzi 53, Milano (Italy)] [Istituto di Fisica del Plasma, Associazione EURATOM-ENEA-CNR, via Roberto Cozzi 53, Milano (Italy); Giacomelli, L.; Rebai, M. [Department of Physics “G. Occhialini,” University of Milano Bicocca, Piazza della Scienza 3, Milano (Italy)] [Department of Physics “G. Occhialini,” University of Milano Bicocca, Piazza della Scienza 3, Milano (Italy); Griesmayer, E. [Atominstitut, Vienna University of Technology (Austria)] [Atominstitut, Vienna University of Technology (Austria); Kaveney, G.; Syme, B. [Culham Centre for Fusion Energy, Culham OX143DB (United Kingdom)] [Culham Centre for Fusion Energy, Culham OX143DB (United Kingdom); Collaboration: JET-EFDA Contributors

    2014-04-15

    First simultaneous measurements of deuterium-deuterium (DD) and deuterium-tritium neutrons from deuterium plasmas using a Single crystal Diamond Detector are presented in this paper. The measurements were performed at JET with a dedicated electronic chain that combined high count rate capabilities and high energy resolution. The deposited energy spectrum from DD neutrons was successfully reproduced by means of Monte Carlo calculations of the detector response function and simulations of neutron emission from the plasma, including background contributions. The reported results are of relevance for the development of compact neutron detectors with spectroscopy capabilities for installation in camera systems of present and future high power fusion experiments.

  15. Conversion electrons used to monitor the energy scale of electron spectrometer near tritium endpoint - a simulation study

    E-Print Network [OSTI]

    M. Rysavy

    2006-01-15

    Measurements of the endpoint region of the tritium beta-decay spectrum provides good possibility to determine neutrino mass. This, however, needs a perfect monitoring of the spectrometer energy scale. A parallel measurement of electron line of known energy - in particular the 83mKr conversion K-line - may serve well to this purpose. The 83Rb decaying to 83mKr seems to be a very suitable radioactive source due to its halflife of 86.2 day. In this work, we determine the amount of 83Rb which is necessary for a successful monitoring.

  16. ARM Climate Research Facility

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity ofkandz-cm11 OutreachProductswsicloudwsicloudden Documentation Data Management Facility PlotsProducts (VAP) VAP3 ARM Assists1

  17. ARM Climate Research Facility

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity ofkandz-cm11 OutreachProductswsicloudwsicloudden Documentation Data Management Facility PlotsProducts (VAP) VAP3 ARM Assists11

  18. ARM Climate Research Facility

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity ofkandz-cm11 OutreachProductswsicloudwsicloudden Documentation Data Management Facility PlotsProducts (VAP) VAP3 ARM Assists113

  19. ARM Climate Research Facility

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity ofkandz-cm11 OutreachProductswsicloudwsicloudden Documentation Data Management Facility PlotsProducts (VAP) VAP3 ARM

  20. ARM Climate Research Facility

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity ofkandz-cm11 OutreachProductswsicloudwsicloudden Documentation Data Management Facility PlotsProducts (VAP) VAP3 ARM8 ARM