National Library of Energy BETA

Sample records for tritium extraction facility

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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)

  6. Type B Investigation Board Report on the April 2, 2002, Worker Fall from Shoring/Scaffolding Structure at the Savannah River Site Tritium Extraction Facility Construction Site

    Office of Energy Efficiency and Renewable Energy (EERE)

    On April 2, 2002, a carpenter helping to erect shoring/scaffolding fell about 52” and struck his head. He sustained head injuries requiring hospitalization that exceeded the threshold for a Type B investigation in accordance with Department of Energy (DOE) Order 225.1A, Accident Investigation. The accident occurred at the DOE’s Savannah River Site (SRS) at the Tritium Extraction Facility (TEF) construction site.

  7. Radiological Training for Tritium Facilities

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

    DOE HANDBOOK RADIOLOGICAL TRAINING FOR TRITIUM FACILITIES U.S. Department of Energy AREA TRNG Washington, D.C. 20585 DISTRIBUTION STATEMENT A. Approved for public release; distribution is unlimited. TS This document has been reproduced from the best available copy. Available to DOE and DOE contractors from ES&H Technical Information Services, U.S. Department of Energy, (800) 473-4375, fax: (301) 903-9823. Available to the public from the U.S. Department of Commerce, Technology

  8. Radiological Training for Tritium Facilities

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

    Change Notice No. 2 May 2007 DOE HANDBOOK RADIOLOGICAL TRAINING FOR TRITIUM FACILITIES U.S. Department of Energy AREA TRNG Washington, D.C. 20585 DISTRIBUTION STATEMENT A. Approved for public release; distribution is unlimited. TS This document has been reproduced from the best available copy. Available to DOE and DOE contractors from ES&H Technical Information Services, U.S. Department of Energy, (800) 473-4375, fax: (301) 903-9823. Available to the public from the U.S. Department of

  9. Independent Oversight Review, Savannah River Site Tritium Facilities- December 2012

    Office of Energy Efficiency and Renewable Energy (EERE)

    Review of Site Preparedness for Severe Natural Phenomena Events at the Savannah River Site Tritium Facilities

  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. Tritium extraction throughput at Hanford, 1949--1954

    SciTech Connect (OSTI)

    Gydesen, S.P.

    1994-02-24

    Two tritium extraction campaigns were conducted at the 108 B facility. Both glass and metal extraction lines were utilized during the first campaign which began in February 1949 and was completed in March 1952. Five glass lines were constructed and made available for use as needed. Operation of the metal extraction line was begun on May 3, 1951. It continued in production until completion of the first campaign in March 1952. The second campaign used only the metal line. It was initiated in December 1953 and fulfilled in August 1954. Tritium production and extraction throughput information from Hanford operations was recently declassified. This document presents tritium extraction throughput information excerpted from monthly production reports which remain classified SECRET-RESTRICTED DATA because they contain information on weapon part fabrication, shipments, tritium technology and unit costs. Individuals with the appropriate level of clearance and need-to-know may request access to these reports through the DOE or appropriate Hanford contractor, following established written procedures. This data was collected for use by the Source Term Task Leader of the hanford Environmental Dose Reconstruction Project, to develop a source term for tritium to meet a 1994 milestone. The extraction quantities for the two campaigns are presented.

  12. Radiological training for tritium facilities

    SciTech Connect (OSTI)

    1996-12-01

    This program management guide describes a recommended implementation standard for core training as outlined in the DOE Radiological Control Manual (RCM). The standard is to assist those individuals, both within DOE and Managing and Operating contractors, identified as having responsibility for implementing the core training recommended by the RCM. This training may also be given to radiological workers using tritium to assist in meeting their job specific training requirements of 10 CFR 835.

  13. Quantitative fire risk assessment for a proposed tritium technology facility

    SciTech Connect (OSTI)

    Zeng, Y. )

    1991-01-01

    A new Tritium Technology Facility has been proposed for the Chalk River Laboratories to support fusion research and the commercial use of tritium. One of the major safety and licensing issues for the new facility raised by the internal Safety Review Committee is the potential hazard fire poses to it. Fire could cause a large release from tritium from the facility's metal tritide storage beds, resulting in conversion of elemental tritium (HT) into oxide tritium (HTO). The radiological hazard of HTO is {approximately}10,000 times higher than that of HT. Because of the potential significance of fire in the tritium facility, a quantitative fire risk assessment has been conducted for the proposed new facility. The frequency of a large tritium release due to a fire in the Tritium Technology Facility was assessed as being on the order of 10{sup {minus}5} per year, which satisfies the safety goal requirement of the facility.

  14. 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.

  15. Software quality assurance at the weapons engineering tritium facility

    SciTech Connect (OSTI)

    Hart, O.

    1997-11-01

    This report contains viewgraphs on the evolution of software quality assurance at the Weapons Engineering Tritium Facility in relation to DOE`s requirements for nuclear facilities.

  16. Report of the Task Group on operation Department of Energy tritium facilities

    SciTech Connect (OSTI)

    Not Available

    1991-10-01

    This report discusses the following topics on the operation of DOE Tritium facilities: Environment, Safety, and Health Aspects of Tritium; Management of Operations and Maintenance Functions; Safe Shutdown of Tritium Facilities; Management of the Facility Safety Envelope; Maintenance of Qualified Tritium Handling Personnel; DOE Tritium Management Strategy; Radiological Control Philosophy; Implementation of DOE Requirements; Management of Tritium Residues; Inconsistent Application of Requirements for Measurement of Tritium Effluents; Interdependence of Tritium Facilities; Technical Communication among Facilities; Incorporation of Confinement Technologies into New Facilities; Operation/Management Requirements for New Tritium Facilities; and Safety Management Issues at Department of Energy Tritium Facilities.

  17. TRITIUM 2013

    Office of Environmental Management (EM)

    www-fusion-magnetique.cea.fr/tritium2013/index.html 1. Containment, safety, and environmental impact 2. Decontamination and waste management 3. Water and air detritiation 4. Tritium processing (purification, isotopic separation,...) 5. Tritium facilities and operation 6. Biological effects 7. Interactions with materials 8. Tracer technics and isotopic effects 9. Measurement, monitoring and accountancy 10. Tritium supply, transport and storage 11. Tritium breeding and extraction 12. Ot

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

    SciTech Connect (OSTI)

    Hsu, R.H.; 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.

  19. Closing the TSTA Facility, tritium removed from TSTA

    SciTech Connect (OSTI)

    Tesch, Charles; Rogers, M. L.; Michelotti, R. A.

    2004-01-01

    The Tritium Systems Test Assembly (TSTA) project was begun in 1978 to develop, design, and demonstrate the technology and safe operation of selected tritium processing systems required for a fusion reactor. The TSTA is located at Los Alamos National Laboratory in Los Alamos, New Mexico, and was initially funded by the US DOE. Tritium processing at TSTA began in 1984. In 2001, DOE determined that the mission of TSTA had been successfully completed, and the facility should be stabilized. Stabilization comprised placing the facility in a safe and stable configuration with a goal of reducing the tritium inventory to below the DOE low-hazard nuclear facility threshold of 16000 Ci. The facility was then to be held in this safe and stable state until funding was available for the final decontamination and decommissioning. This paper will describe the process and results of the activities required to achieve the safe and stable condition. At the completion of the TSTA mission, the tritium inventory at TSTA was 170 grams. The facility was categorized as a DOE moderate-hazard nuclear facility. At the completion of the stabilization project in 2003, the tritium inventory had been reduced to less than 1 gram, well below the low-hazard nuclear facility threshold, and the facility was categorized as a radiological facility. The pre-stabilization tritium inventory at TSTA was grouped in the following categories: tritium gas mixed with hydrogen isotopes, tritiated water absorbed on molecular sieve, tritium held up as a hydride on various metals, and tritium held up in process components. For each category, the tritium content was characterized, a path for removal was determined, and the proper disposal package was developed. Half of the tritium removed from the facility was reusable and the other half was disposed as waste. Hydrogen exchange, calorimetry, direct sampling, pressure/composition/temperature, radiological smear surveys, and controlled regeneration were methods used to

  20. Development of a tritium recovery system from CANDU tritium removal facility

    SciTech Connect (OSTI)

    Draghia, M.; Pasca, G.; Porcariu, F.

    2015-03-15

    The main purpose of the Tritium Recovery System (TRS) is to reduce to a maximum possible extent the release of tritium from the facility following a tritium release in confinement boundaries and also to have provisions to recover both elemental and vapors tritium from the purging gases during maintenance and components replacement from various systems processing tritium. This work/paper proposes a configuration of Tritium Recovery System wherein elemental tritium and water vapors are recovered in a separated, parallel manner. The proposed TRS configuration is a combination of permeators, a platinum microreactor (MR) and a trickle bed reactor (TBR) and consists of two branches: one branch for elemental tritium recovery from tritiated deuterium gas and the second one for tritium recovery from streams containing a significant amount of water vapours but a low amount, below 5%, of tritiated gas. The two branches shall work in a complementary manner in such a way that the bleed stream from the permeators shall be further processed in the MR and TBR in view of achieving the required decontamination level. A preliminary evaluation of the proposed TRS in comparison with state of the art tritium recovery system from tritium processing facilities is also discussed. (authors)

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

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

    Savannah River Ecology Laboratory Tritium Irrigation Facility and Automated Vadose Monitoring System The opportunity to study tritium movement in a natural system presents a rare opportunity for both physical and biological research. Researchers may take advantage of tritium's properties as a conservative tracer for modeling contaminant transport, as a radioactive tracer for examining biological processes involving water, or as an example of radionuclide contaminant behavior in natural

  2. Independent Oversight Review, Savannah River Field Office Tritium Facilities – November 2013

    Broader source: Energy.gov [DOE]

    Review of Savannah River Field Office Tritium Facilities Radiological Controls Activity-Level Implementation

  3. Measurement and Control Systems of Tritium Facilities for Scientific Research

    SciTech Connect (OSTI)

    Vinogradov, Yu.I.; Kuryakin, A.V.; Yukhimchuk, A.A.

    2005-07-15

    The technical approach, equipment and software developed during the creation of measurement and control systems for two complexes are described. The first one is a complex that prepares the gas mixture and targets of the 'TRITON' facility. The 'TRITON' facility is designed for studying muon catalyzed fusion reactions in triple mixtures of H/D/T hydrogen isotopes over wide ranges of temperature and pressure. The second one is 'ACCULINNA' - the liquid tritium target designed to investigate the neutron overloaded hydrogen and helium nuclei. These neutron-overloaded nuclei are produced in reactions of tritium beams on a heavy hydrogen and tritium target.

  4. 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.

  5. Use of the fast flux test facility for tritium production

    SciTech Connect (OSTI)

    Drell, S.; Hammer, D.; Cornwall, J.M.; Dyson, F.; Garwin, R.

    1996-10-25

    This report provides the results of a JASON review of the technical feasibility of using the Department of Energy`s (DOE`s) Fast Flux Test Facility (FFTF) to generate tritium needed for the enduring United States nuclear weapons stockpile.

  6. Zeolite membrane cascade for tritium extraction and recovery systems

    SciTech Connect (OSTI)

    Borisevich, O.; Demange, D.; Lefebvre, X.; Kind, M.

    2015-03-15

    Membrane separation by zeolite membranes has been proposed as a pre-concentration stage for the tritium extraction from the purge helium of the breeding blanket combined with a final recovery by the catalytic membrane reactor PERMCAT. This fully continuous operation improves the tritium management in fusion machines, minimizing the tritium inventory. For the first time, the permeation measurements for H{sub 2} - He mixtures through a MFI-alumina hollow fibre membrane has been measured for different compositions (0.1 - 20% H{sub 2}) and temperatures. Such a highly permeable membrane, although it shows a limited selectivity, appears attractive for tritium recovery in the blanket. This will imply its operation in a membrane cascade, for which simulation work is ongoing. Mathematically the process is modeled using mass balance equations that can be transformed into the matrix form and solved iteratively assuming a permeate concentration on the first step of iteration, until the separation requirements are fulfilled.

  7. Health physics manual of good practices for tritium facilities

    SciTech Connect (OSTI)

    Blauvelt, R.K.; Deaton, M.R.; Gill, J.T.

    1991-12-01

    The purpose of this document is to provide written guidance defining the generally accepted good practices in use at Department of Energy (DOE) tritium facilities. A {open_quotes}good practice{close_quotes} is an action, policy, or procedure that enhances the radiation protection program at a DOE site. The information selected for inclusion in this document should help readers achieve an understanding of the key radiation protection issues at tritium facilities and provide guidance as to what characterizes excellence from a radiation protection point of view. The ALARA (As Low as Reasonable Achievable) program at DOE sites should be based, in part, on following the good practices that apply to their operations.

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

    Broader source: Energy.gov [DOE]

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

  9. 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.

  10. NPH Risk Assessment and Mitigation of a SRS Facility for the Safe Storage of Tritium

    SciTech Connect (OSTI)

    Joshi, J.R.; Griffin, M.J.; Bjorkman, G.S.

    1995-10-18

    Because of the reduction in the nation`s stockpile of weapon systems a large amount of tritium is being returned to the Savannah River Site in Aiken, SC. Due to the increased quantity of tritium returning to SRS, the SRS Tritium Facility was tasked to determine the most cost effective means to safely store the tritium gas in a short period of time. This paper presents results of the risk assessment developed to evaluate the safe storage of tritium at SRS, and highlights the structural design of the HIVES used as the cost-effective short term NPH mitigation solution.

  11. Material Control & Accountability for Department Of Energy (DOE) 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.

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

    Office of Environmental Management (EM)

    Tritium Permeation Activity at Safety and Tritium Applied Research (STAR) Facility Tritium Plasma Experiment and Its Role in PHENIX Program Meeting Attendance - 33rd Tritium Focus ...

  13. Electrolytic Tritium Extraction in Molten Li-LiT

    Broader source: Energy.gov [DOE]

    Presentation from the 36th Tritium Focus Group Meeting held in Los Alamos, New Mexico, November 3-5, 2015.

  14. 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.

  15. Vacuum Permeator Analysis for Extraction of Tritium from DCLL Blankets

    SciTech Connect (OSTI)

    Humrickhouse, Paul Weston; Merrill, Brad Johnson

    2014-11-01

    It is envisioned that tritium will be extracted from DCLL blankets using a vacuum permeator. We derive here an analytical solution for the extraction efficiency of a permeator tube, which is a function of only two dimensionless numbers: one that indicates whether radial transport is limited in the PbLi or in the solid membrane, and another that is the ratio of axial and radial transport times in the PbLi. The permeator efficiency is maximized by decreasing the velocity and tube diameter, and increasing the tube length. This is true regardless of the mass transport correlation used; we review several here and find that they differ little, and the choice of correlation is not a source of significant uncertainty here. The PbLi solubility, on the other hand, is a large source of uncertainty, and we identify upper and lower bounds from the literature data. Under the most optimistic assumptions, we find that a ferritic steel permeator operating at 550 °C will need to be at least an order of magnitude larger in volume than previous conceptual designs using niobium and operating at higher temperatures.

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

    National Nuclear Security Administration (NNSA)

    and decommissioning of several 1950s era structures. Tritium is a heavy isotope of hydrogen and a key component of nuclear weapons, but it decays radioactively at the rate of...

  17. REPORT OF SURVEY OF THE LOS ALAMOS TRITIUM SYSTEMS TEST ASSEMBLY FACILITY

    Office of Environmental Management (EM)

    REPORT OF SURVEY OF THE LOS ALAMOS TRITIUM SYSTEMS TEST ASSEMBLY FACILITY U.S. Department of Energy Office of Environmental Management & Office of Science Report of Survey of the Los Alamos Tritium Systems Test Assembly Facility Rev. E (Final) October 3, 2000 Contents 1. Introduction 1.1 Purpose 1.2 Facility Description 1.3 Organization Representatives 1.4 Survey Participants 2. Summary, Conclusions & Recommendations 2.1 Comparison With LCAM Requirements 2.2 Transfer Considerations 2.3

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

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

    Facilities - June 2012 More Documents & Publications Technical Qualification Program Self-Assessment Report - Savannah River Site Office - 2011 Independent Oversight Review,...

  19. 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.

  20. Microsoft PowerPoint - Tritium Gas Stream Scrubbing using In-situ Reactive Materials.pptx

    Office of Environmental Management (EM)

    Stream Scrubbing using In-situ Reactive Materials Paul Korinko, Simona Murph, and George Larsen Tritium Focus Group Meeting LANL Nov 3-5, 2015 SRNL-STI-2015-00597 Tritium Production and Extraction * Tritium Producing Burnable Absorber Rods (TPBARs) * Built to strict materials specifications * Coatings, ceramics, metals, processes * Meet NQA-1 requirements * Irradiated in a commercial light water reactor * Extracted at SRS in the Tritium Extraction Facility * Waste disposed on-site Contamination

  1. Development of a tritium monitor combined with an electrochemical tritium pump using a proton conducting oxide

    SciTech Connect (OSTI)

    Tanaka, M.; Sugiyama, T.

    2015-03-15

    The detection of low level tritium is one of the key issues for tritium management in tritium handling facilities. Such a detection can be performed by tritium monitors based on proton conducting oxide technique. We tested a tritium monitoring system composed of a commercial proportional counter combined with an electrochemical hydrogen pump equipped with CaZr{sub 0.9}In{sub 0.1}O{sub 3-α} as proton conducting oxide. The hydrogen pump operated at 973 K under electrolysis conditions using tritiated water vapor (HTO). The proton conducting oxide extracts tritium molecules (HT) from HTO and tritium concentration is measured by the proportional counter. The advantage of the proposed tritium monitoring system is that it is able to convert HTO into molecular hydrogen.

  2. Tritium Focus Group Meeting:

    Office of Environmental Management (EM)

    32 nd Tritium Focus Group Meeting: Tritium research activities in Safety and Tritium Applied Research (STAR) facility, Idaho National Laboratory Masashi Shimada Fusion Safety Program, Idaho National Laboratory April 25 th 2013, Germantown, MD STI #: INL/MIS-13-28975 Outlines 1. Motivation of tritium research activity in STAR facility 2. Unique capabilities in STAR facility 3. Research highlights from tritium retention in HFIR neutron- irradiated tungsten April 25th 2013 Germantown, MD STAR

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

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

    Superseding DOE-HDBK-1105-96 December 1996 DOE HANDBOOK RADIOLOGICAL TRAINING FOR TRITIUM FACILITIES U.S. Department of Energy AREA TRNG Washington, D.C. 20585 DISTRIBUTION STATEMENT A. Approved for public release; distribution is unlimited. TS This document has been reproduced from the best available copy. Available to DOE and DOE contractors from ES&H Technical Information Services, U.S. Department of Energy, (800) 473-4375, fax: (301) 903-9823. Available to the public from the U.S.

  4. Construction and Operation of a Tritium Extraction Facility at...

    National Nuclear Security Administration (NNSA)

    ... NESHAP National Emissions Standards for Hazardous Air Pollutants NFPA National Fire ...EIS-0271 Acronyms and Abbreviations March 1999 xii O 3 Ozone OSHA Occupational Safety and ...

  5. Tritium monitor

    DOE Patents [OSTI]

    Chastagner, Philippe

    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.

  6. 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.

  7. Evaluation of medical isotope production with the accelerator production of tritium (APT) facility

    SciTech Connect (OSTI)

    Benjamin, R.W.; Frey, G.D.; McLean, D.C., Jr; Spicer, K.M.; Davis, S.E.; Baron, S.; Frysinger, J.R.; Blanpied, G.; Adcock, D.

    1997-07-10

    The accelerator production of tritium (APT) facility, with its high beam current and high beam energy, would be an ideal supplier of radioisotopes for medical research, imaging, and therapy. By-product radioisotopes will be produced in the APT window and target cooling systems and in the tungsten target through spallation, neutron, and proton interactions. High intensity proton fluxes are potentially available at three different energies for the production of proton- rich radioisotopes. Isotope production targets can be inserted into the blanket for production of neutron-rich isotopes. Currently, the major production sources of radioisotopes are either aging or abroad, or both. The use of radionuclides in nuclear medicine is growing and changing, both in terms of the number of nuclear medicine procedures being performed and in the rapidly expanding range of procedures and radioisotopes used. A large and varied demand is forecast, and the APT would be an ideal facility to satisfy that demand.

  8. 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.

  9. Independent Oversight Review of the Savannah River Field Office Tritium Facilities Radiological Controls Activity-Level Implementation, November 2013

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

    Independent Oversight Review of the Savannah River Field Office Tritium Facilities Radiological Controls Activity-Level Implementation May 2011 November 2013 Office of Safety and Emergency Management Evaluations Office of Enforcement and Oversight Office of Health, Safety and Security U.S. Department of Energy Table of Contents 1.0 Purpose................................................................................................................................................ 1 2.0

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

    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.

  11. Tritium Permeation Activity at Safety and Tritium Applied

    Office of Environmental Management (EM)

    Permeation Activity at Safety and Tritium Applied Research (STAR) facility Masashi Shimada and Bob Pawelko Fusion Safety Program Idaho National Laboratory Tritium Focus Group meeting September 23-25, 2014 at Idaho National Laboratory, Idaho Falls, ID Outline: 1. Motivation of low tritium partial pressure permeation 2. Tritium permeation for fission application 3. Tritium permeation for fusion application M.Shimada | Tritium Focus Group meeting | Idaho Falls, ID | September 23-25, 2014 2

  12. Tritium Technology at CNL

    Office of Environmental Management (EM)

    1- UNRESTRICTED/ ILLIMITÉ Chalk River Tritium Activities: Select Topics Presented by Hugh Boniface Tritium Focus Group Meeting, Princeton, NJ, 2015 May -2- UNRESTRICTED/ ILLIMITÉ * Canadian Nuclear Labs - former AECL * New Tritium Facility * Tritium-resistant e-cell materials * Beta-voltaics * Helium-3 recovery Topics -3- UNRESTRICTED/ ILLIMITÉ * Main campus of Canadian Nuclear Labs - former AECL * Established 1952 Crown Corporation * 3100 employees (500 advanced degrees) * 600 M$ in revenue

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

    National Nuclear Security Administration (NNSA)

    ... production reactor and a nuclear power plant without tritium production are as ... from the CLWR site to the Savannah River Site for tritium extraction and purification. ...

  14. Tritium Detection Methods and Limitations

    Office of Environmental Management (EM)

    Detection Methods and Limitations Tritium Focus Group Meeting, April 2014 Tom Voss, Northern New Mexico DOE-HDBK-1105-2002 RADIOLOGICAL TRAINING FOR TRITIUM FACILITIES U.S. Department of Energy AREA TRNG Washington, D.C. 20585 DISTRIBUTION STATEMENT A. Approved for public release; distribution is unlimited. DOE-HDBK-1105-2002 Radiological Training for Tritium Facilities U.S. Department of Energy, Radiological Control Programs for Special Tritium Compounds, DOE-STD- draft, Washington, D.C.

  15. 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 Presentation * Introduction & Background * Tritium Facilities: Laboratories, old and new * Tritium Separations: CECE process * Tritium Properties: Materials * Tritium Exploitation: Batteries, Helium-3 * Other work: Education, environment, biology, fusion, hydrogen UNRESTRICTED / ILLIMITÉ 2 Background UNRESTRICTED /

  16. Tritium Operation Improvements at the Idaho National Laboratory (INL)

    Office of Environmental Management (EM)

    Safety and Tritium Applied Research (STAR) facility | Department of Energy Operation Improvements at the Idaho National Laboratory (INL) Safety and Tritium Applied Research (STAR) facility Tritium Operation Improvements at the Idaho National Laboratory (INL) Safety and Tritium Applied Research (STAR) facility Presentation from the 35th Tritium Focus Group Meeting held in Princeton, New Jersey on May 05-07, 2015. Tritium Operation Improvements at the INL STAR facility (3.33 MB) More Documents

  17. Facility effluent monitoring plan for the plutonium uranium extraction facility

    SciTech Connect (OSTI)

    Wiegand, D.L.

    1994-09-01

    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 they are adequate to ensure the public health and safety as specified in applicable federal, state, and local requirements. This facility effluent monitoring plan shall 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 of every three years.

  18. 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.

  19. Transportation risk assessment for the shipment of irradiated FFTF tritium target assemblies from the Hanford Site to the Savannah River Site

    SciTech Connect (OSTI)

    Nielsen, D. L.

    1997-11-19

    A Draft Technical Information Document (HNF-1855) is being prepared to evaluate proposed interim tritium and medical isotope production at the Fast Flux Test Facility (FFTF). This report examines the potential health and safety impacts associated with transportation of irradiated tritium targets from FFTF to the Savannah River Site for processing at the Tritium Extraction Facility. Potential risks to workers and members of the public during normal transportation and accident conditions are assessed.

  20. 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.

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

    SciTech Connect (OSTI)

    Dppner, 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.; Dittrich, T. R.; Haan, S.; Kritcher, A. L.; MacPhee, A.; Le Pape, S.; Pak, A.; Patel, P. K.; Springer, P. T.; Salmonson, J. D.; Tommasini, R.; Benedetti, L. R.; Bond, E.; Bradley, D. K.; Caggiano, J.; Church, J.; Dixit, S.; Edgell, D.; Edwards, M. J.; Fittinghoff, D. N.; Frenje, J.; Gatu Johnson, M.; Grim, G.; Hatarik, R.; Havre, M.; Herrmann, H.; Izumi, N.; Khan, S. F.; Kline, J. L.; Knauer, J.; Kyrala, G. A.; Landen, O. L.; Merrill, F. E.; Moody, J.; Moore, A. S.; Nikroo, A.; Ralph, J. E.; Remington, B. A.; Robey, H.; Sayre, D.; Schneider, M.; Streckert, H.; Town, R.; Turnbull, D.; Volegov, P. L.; Wan, A.; Widmann, K.; Wilde, C. H.; Yeamans, C.

    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 shape 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.

  2. 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

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

    SciTech Connect (OSTI)

    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.; Dittrich, T. R.; Haan, S.; Kritcher, A. L.; MacPhee, A.; Le Pape, S.; Pak, A.; Patel, P. K.; Springer, P. T.; Salmonson, J. D.; Tommasini, R.; Benedetti, L. R.; Bond, E.; Bradley, D. K.; Caggiano, J.; Church, J.; Dixit, S.; Edgell, D.; Edwards, M. J.; Fittinghoff, D. N.; Frenje, J.; Gatu Johnson, M.; Grim, G.; Hatarik, R.; Havre, M.; Herrmann, H.; Izumi, N.; Khan, S. F.; Kline, J. L.; Knauer, J.; Kyrala, G. A.; Landen, O. L.; Merrill, F. E.; Moody, J.; Moore, A. S.; Nikroo, A.; Ralph, J. E.; Remington, B. A.; Robey, H.; Sayre, D.; Schneider, M.; Streckert, H.; Town, R.; Turnbull, D.; Volegov, P. L.; Wan, A.; Widmann, K.; Wilde, C. H.; Yeamans, C.

    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 shape 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.

  4. 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.

  5. Potential role of the Fast Flux Test Facility and the advanced test reactor in the U.S. tritium production system

    SciTech Connect (OSTI)

    Dautel, W.A.

    1996-10-01

    The Deparunent of Energy is currently engaged in a dual-track strategy to develop an accelerator and a conunercial light water reactor (CLWR) as potential sources of tritium supply. New analysis of the production capabilities of the Fast Flux Test Facility (FFTF) at the Hanford Site argues for considering its inclusion in the tritium supply,system. The use of the FFTF (alone or together with the Advanced Test Reactor [ATR] at the Idaho National Engineering Laboratory) as an integral part of,a tritium production system would help (1) ensure supply by 2005, (2) provide additional time to resolve institutional and technical issues associated with the- dual-track strategy, and (3) reduce discounted total life-cycle`costs and near-tenn annual expenditures for accelerator-based systems. The FFRF would also provide a way to get an early start.on dispositioning surplus weapons-usable plutonium as well as provide a source of medical isotopes. Challenges Associated With the Dual-Track Strategy The Departinent`s purchase of either a commercial reactor or reactor irradiation services faces challenging institutional issues associated with converting civilian reactors to defense uses. In addition, while the technical capabilities of the individual components of the accelerator have been proven, the entire system needs to be demonstrated and scaled upward to ensure that the components work toge ther 1548 as a complete production system. These challenges create uncertainty over the ability of the du2a-track strategy to provide an assured tritium supply source by 2005. Because the earliest the accelerator could come on line is 2007, it would have to operate at maximum capacity for the first few years to regenerate the reserves lost through radioactive decay aftei 2005.

  6. Performance of Vacuum Pumps to be Used in Tritium Extraction Facility

    SciTech Connect (OSTI)

    Steimke, J.L.

    1999-04-06

    The goal of this test was to measure pump operating characteristics for three different gases and a wider range of conditions than for the vendor data. Test results will be used by Engineering Development Section for incorporation in a computer model of the pump train.

  7. Tritium 2016

    Office of Environmental Management (EM)

    2016 11 TH International Conference on Tritium Science and Technology Robert Addis Director, Defense Programs Technology, SRNL Tritium Focus Group Meeting. Savannah River Site, Aiken, SC 29808 April 22, 2014 History of International Conference on Tritium Science and Technology Publication Pub Date * 1 st 1980 Dayton, USA * 2 nd 1985 Dayton, USA Fusion Tech 8, 2 (1985) * 3 rd 1988 Toronto, Canada Fusion Tech 14, 2 (1988) * 4 th 1991 Albuquerque, USA Fusion Tech 21, 2 (1992) * 5 th 1995 Belgirate,

  8. 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.

  9. Evaluation of tritium release properties of advanced tritium breeders

    SciTech Connect (OSTI)

    Hoshino, T.; Ochiai, K.; Edao, Y.; Kawamura, Y.

    2015-03-15

    Demonstration power plant (DEMO) fusion reactors require advanced tritium breeders with high thermal stability. Lithium titanate (Li{sub 2}TiO{sub 3}) advanced tritium breeders with excess Li (Li{sub 2+x}TiO{sub 3+y}) are stable in a reducing atmosphere at high temperatures. Although the tritium release properties of tritium breeders are documented in databases for DEMO blanket design, no in situ examination under fusion neutron (DT neutron) irradiation has been performed. In this study, a preliminary examination of the tritium release properties of advanced tritium breeders was performed, and DT neutron irradiation experiments were performed at the fusion neutronics source (FNS) facility in JAEA. Considering the tritium release characteristics, the optimum grain size after sintering is <5 μm. From the results of the optimization of granulation conditions, prototype Li{sub 2+x}TiO{sub 3+y} pebbles with optimum grain size (<5 μm) were successfully fabricated. The Li{sub 2+x}TiO{sub 3+y} pebbles exhibited good tritium release properties similar to the Li{sub 2}TiO{sub 3} pebbles. In particular, the released amount of HT gas for easier tritium handling was higher than that of HTO water. (authors)

  10. 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.

  11. Tritium Instrument Demonstration Station (TIDS)

    Office of Environmental Management (EM)

    Cortés Concepción, Laura Tovo April 22, 2014 Tritium Focus Group Meeting SRNL-STI-2014-00172 What is the challenge? Tritium Facilities is critically reliant on dated analytical technologies Low-mass, high-resolution mass spectrometer issues: * Near end-of-life (30+ years old) * Spare parts not available from vendor * Vendor support is difficult or unavailable Need for alternative, accessible analytical technologies within DP for: * Complement current analytical methods * Greater ability to

  12. 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.

  13. Tritium Plasma Experiment and Its Role in PHENIX Program | Department of

    Office of Environmental Management (EM)

    Energy Plasma Experiment and Its Role in PHENIX Program Tritium Plasma Experiment and Its Role in PHENIX Program Presentation from the 34th Tritium Focus Group Meeting held in Idaho Falls, Idaho on September 23-25, 2014. Tritium Plasma Experiment and Its Role in PHENIX Program (13.13 MB) More Documents & Publications Tritium Permeation Activity at Safety and Tritium Applied Research (STAR) Facility Fusion Nuclear Science and Technology Program - Status and Plans for Tritium Research

  14. Report of Survey of the Los Alamos Tritium Systems Test Assembly...

    Office of Environmental Management (EM)

    the Los Alamos Tritium Systems Test Assembly Facility Report of Survey of the Los Alamos Tritium Systems Test Assembly Facility The purpose of this document is to report the ...

  15. 32nd Tritium Focus Group Meeting, Cloverleaf Building, Germantown...

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

    in Safety and Tritium Applied Research (STAR) facility, Idaho National ... Hazard Categorization) Methodology Advances in Design of the Next Generation Hydride Bed ...

  16. Disposal of tritium-exposed metal hydrides

    SciTech Connect (OSTI)

    Nobile, A.; Motyka, T.

    1991-01-01

    A plan has been established for disposal of tritium-exposed metal hydrides used in Savannah River Site (SRS) tritium production or Materials Test Facility (MTF) R D operations. The recommended plan assumes that the first tritium-exposed metal hydrides will be disposed of after startup of the Solid Waste Disposal Facility (SWDF) Expansion Project in 1992, and thus the plan is consistent with the new disposal requiremkents that will be in effect for the SWDF Expansion Project. Process beds containing tritium-exposed metal hydride powder will be disposed of without removal of the powder from the bed; however, disposal of tritium-exposed metal hydride powder that has been removed from its process vessel is also addressed.

  17. Disposal of tritium-exposed metal hydrides

    SciTech Connect (OSTI)

    Nobile, A.; Motyka, T.

    1991-12-31

    A plan has been established for disposal of tritium-exposed metal hydrides used in Savannah River Site (SRS) tritium production or Materials Test Facility (MTF) R&D operations. The recommended plan assumes that the first tritium-exposed metal hydrides will be disposed of after startup of the Solid Waste Disposal Facility (SWDF) Expansion Project in 1992, and thus the plan is consistent with the new disposal requiremkents that will be in effect for the SWDF Expansion Project. Process beds containing tritium-exposed metal hydride powder will be disposed of without removal of the powder from the bed; however, disposal of tritium-exposed metal hydride powder that has been removed from its process vessel is also addressed.

  18. Experiences with decontaminating tritium-handling apparatus

    SciTech Connect (OSTI)

    Maienschein, J.L.; Garcia, F.; Garza, R.G.; Kanna, R.L.; Mayhugh, S.R.; Taylor, D.T. )

    1992-03-01

    Tritium-handling apparatus has been decontaminated as part of the downsizing of the LLNL Tritium Facility. Two stainless-steel glove boxes that had been used to process lithium deuteride-tritide (LiDT) slat were decontaminated using the Portable Cleanup System so that they could be flushed with room air through the facility ventilation system. In this paper the details on the decontamination operation are provided. A series of metal (palladium and vanadium) hydride storage beds have been drained of tritium and flushed with deuterium, in order to remove as much tritium as possible. The bed draining and flushing procedure is described, and a calculational method is presented which allows estimation of the tritium remaining in a bed after it has been drained and flushed. Data on specific bed draining and flushing are given.

  19. 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.

  20. TRITIUM SESSIONS AT THE

    Office of Environmental Management (EM)

    TRITIUM IN FISSION AND FUSION sponsored by IRD; cosponsored by BMD Session Organizer and Chair: Tom Voss (Cybermesa) Discussion of Tritium Safety in Fusion Reactors, Satoshi ...

  1. 1997 evaluation of tritium removal and mitigation technologies for Hanford Site wastewaters

    SciTech Connect (OSTI)

    Jeppson, D.W.; Biyani, R.K.; Duncan, J.B.; Flyckt, D.L.; Mohondro, P.C.; Sinton, G.L.

    1997-07-24

    This report contains results of a biennial assessment of tritium separation technology and tritium nitration techniques for control of tritium bearing wastewaters at the Hanford Site. Tritium in wastewaters at Hanford have resulted from plutonium production, fuel reprocessing, and waste handling operations since 1944. this assessment was conducted in response to the Hanford Federal Facility Agreement and Consent Order.

  2. 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...

  3. Tritium glovebox stripper system seismic design evaluation

    SciTech Connect (OSTI)

    Grinnell, J. J.; Klein, J. E.

    2015-09-01

    The use of glovebox confinement at US Department of Energy (DOE) tritium facilities has been discussed in numerous publications. Glovebox confinement protects the workers from radioactive material (especially tritium oxide), provides an inert atmosphere for prevention of flammable gas mixtures and deflagrations, and allows recovery of tritium released from the process into the glovebox when a glovebox stripper system (GBSS) is part of the design. Tritium recovery from the glovebox atmosphere reduces emissions from the facility and the radiological dose to the public. Location of US DOE defense programs facilities away from public boundaries also aids in reducing radiological doses to the public. This is a study based upon design concepts to identify issues and considerations for design of a Seismic GBSS. Safety requirements and analysis should be considered preliminary. Safety requirements for design of GBSS should be developed and finalized as a part of the final design process.

  4. Continuous aqueous tritium monitor

    DOE Patents [OSTI]

    McManus, Gary J. (Idaho Falls, ID); Weesner, Forrest J. (Idaho Falls, ID)

    1989-05-30

    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 and 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.

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

    Office of Environmental Management (EM)

    (7.98 MB) More Documents & Publications Tritium Behavior in Lead Lithium Eutectic (LLE) at Low Tritium Partial Pressure Tritium Plasma Experiment and Its Role in PHENIX Program

  6. Magmatic tritium

    SciTech Connect (OSTI)

    Goff, F.; Aams, A.I.; McMurtry, G.M.; Shevenell, L.; Pettit, D.R.; Stimac, J.A.; Werner, C.

    1997-07-01

    This is the final report of a three-year, Laboratory-Directed Research and Development (LDRD) project at the Los Alamos National Laboratory. Detailed geochemical sampling of high-temperature fumaroles, background water, and fresh magmatic products from 14 active volcanoes reveal that they do not produce measurable amounts of tritium ({sup 3}H) of deep origin (<0.1 T.U. or <0.32 pCi/kg H{sub 2}O). On the other hand, all volcanoes produce mixtures of meteoric and magmatic fluids that contain measurable {sup 3}H from the meteoric end-member. The results show that cold fusion is probably not a significant deep earth process but the samples and data have wide application to a host of other volcanological topics.

  7. 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.

  8. Facility effluent monitoring plan for the plutonium-uranium extraction facility

    SciTech Connect (OSTI)

    Lohrasbi, J.; Johnson, D.L.; De Lorenzo, D.S.

    1993-12-01

    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 they are adequate to ensure the public health and safety as specified in applicable federal, state, and local requirements. This facility effluent monitoring plan shall 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 of every three years.

  9. EXTRACTION OF FRACTURE-MECHANICS AND TRANSMISSION-ELECTRON-MICROSCOPY SAMPLES FROM TRITIUM-EXPOSED RESERVOIRS USING ELECTRIC-DISCHARGE MACHINING

    SciTech Connect (OSTI)

    Morgan, M; Ken Imrich, K; Michael Tosten, M

    2006-08-31

    The Enhanced Surveillance Campaign is funding a program to investigate tritium aging effects on the structural properties of tritium reservoir steels. The program is designed to investigate how the structural properties of reservoir steels change during tritium service and to examine the role of microstructure and reservoir manufacturing on tritium compatibility. New surveillance tests are also being developed that can better gauge the long-term effects of tritium and its radioactive decay product, helium-3, on the properties of reservoir steels. In order to conduct these investigations, three types of samples are needed from returned reservoirs: tensile, fracture mechanics, and transmission-electron microscopy (TEM). An earlier report demonstrated how the electric-discharge machining (EDM) technique can be used for cutting tensile samples from serial sections of a 3T reservoir and how yield strength, ultimate strength and elongation could be measured from those samples. In this report, EDM was used successfully to section sub-sized fracture-mechanics samples from the inner and outer walls of a 3T reservoir and TEM samples from serial sections of a 1M reservoir. This report fulfills the requirements for the FY06 Level 3 milestone, TSR 15.1 ''Cut Fracture-Mechanics Samples from Tritium-Exposed Reservoir'' and TSR 15.2 ''Cut Transmission-electron-microscopy foils from Tritium-Exposed Reservoir'' for the Enhance Surveillance Campaign (ESC). This was in support of ESC L2-1870 Milestone-''Provide aging and lifetime assessments of selected components and materials for multiple enduring stockpile systems''.

  10. Tritium handling experience at Atomic Energy of Canada Limited

    SciTech Connect (OSTI)

    Suppiah, S.; McCrimmon, K.; Lalonde, S.; Ryland, D.; Boniface, H.; Muirhead, C.; Castillo, I.

    2015-03-15

    Canada has been a leader in tritium handling technologies as a result of the successful CANDU reactor technology used for power production. Over the last 50 to 60 years, capabilities have been established in tritium handling and tritium management in CANDU stations, tritium removal processes for heavy and light water, tritium measurement and monitoring, and understanding the effects of tritium on the environment. This paper outlines details of tritium-related work currently being carried out at Atomic Energy of Canada Limited (AECL). It concerns the CECE (Combined Electrolysis and Catalytic Exchange) process for detritiation, tritium-compatible electrolysers, tritium permeation studies, and tritium powered batteries. It is worth noting that AECL offers a Tritium Safe-Handling Course to national and international participants, the course is a mixture of classroom sessions and hands-on practical exercises. The expertise and facilities available at AECL is ready to address technological needs of nuclear fusion and next-generation nuclear fission reactors related to tritium handling and related issues.

  11. 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.

  12. Tritium Focus Group

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

    matters related to tritium. Contacts Mike Rogers (505) 665-2513 Email Chandra Savage Marsden (505) 664-0183 Email The Tritium Focus Group consists of participants from member...

  13. 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.

  14. 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.

  15. Fermilab | Tritium at Fermilab

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

    Feature photo Tritium is a weakly radioactive form of hydrogen. In nature, it's formed when cosmic particles hit Earth's atmosphere. Here at Fermilab, tritium is an expected byproduct of the operation of our particle accelerators. This website provides information on the monitoring and management of tritium at Fermilab. Small but detectable levels of tritium - well below regulatory limits - are found in some ponds on the Fermilab site, in Indian Creek as it leaves the laboratory and in some of

  16. 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.

  17. 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.

  18. TRITIUM SESSIONS AT THE

    Office of Environmental Management (EM)

    SESSIONS AT THE 2012 ANS MEETING James T (Tom) Voss April 24, 2013 Presented at the 2013 U. S. DOE Tritium Focus Group Meeting TRITIUM IN FISSION AND FUSION sponsored by IRD; cosponsored by BMD Session Organizer and Chair: Tom Voss (Cybermesa) Discussion of Tritium Safety in Fusion Reactors, Satoshi Fukada (Kyushu Univ) Commercial Light Water Production of Tritium: Update and Path Forward, Cheryl K. Thornhill (PNNL) Neutronics Experiments for the European ITER Test Blanket Modules, A. Klix -

  19. 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.

  20. Development of tritium technologies at KAERI

    SciTech Connect (OSTI)

    Chung, H.; Koo, D.; Lee, J.; Park, J.; Yim, S.P.; Yoon, C.; Lim, J.; Choi, W.; Ahn, H.; Kang, H.; Kim, I.; Paek, S.; Yunn, S.H.; Jung, K.J.

    2015-03-15

    Korea has been operating a CANDU nuclear power plant since 1983. Tritium generated in the heavy water of the plant is removed by the Wolsong TRF (Tritium Removal Facility) and measurement campaigns of tritium near the power plant have shown the efficiency of the TRF system. The HANARO reactor uses heavy water as both reflector and moderator. In HANARO the tritiated water removal system consists of compressors, condensers, and adsorption beds. A tritium behavior analysis code (TRIBAC) for a Very High Temperature Gas-Cooled Reactor (VHTR) is under development at KAERI. The TRIBAC computer software has been equipped with models for tritium production, purification, and leakage, as well as chemisorption and tritium behavior, in the hydrogen production system. Korea takes part into the ITER program and is responsible for the supply of an SDS (Tritium Storage and Delivery System). Within this program Korea has launched an experimental program to study the physico-chemical properties of metal and their hydrides in which hydrogen isotope gases can be stored and removed safely.

  1. PDRD (SR13046) TRITIUM PRODUCTION FINAL REPORT

    SciTech Connect (OSTI)

    Smith, P.; Sheetz, S.

    2013-09-30

    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.

  2. Process for recovering tritium from molten lithium metal

    DOE Patents [OSTI]

    Maroni, Victor A.

    1976-01-01

    Lithium tritide (LiT) is extracted from molten lithium metal that has been exposed to neutron irradiation for breeding tritium within a thermonuclear or fission reactor. The extraction is performed by intimately contacting the molten lithium metal with a molten lithium salt, for instance, lithium chloride - potassium chloride eutectic to distribute LiT between the salt and metal phases. The extracted tritium is recovered in gaseous form from the molten salt phase by a subsequent electrolytic or oxidation step.

  3. Tritium Plasma Experiment and

    Office of Environmental Management (EM)

    Plasma Experiment and its role in PHENIX program Masashi Shimada, Chase Taylor Fusion Safety Program Idaho National Laboratory Rob Kolasinski Sandia National Laboratories, Livermore Tritium Focus Group meeting September 23-25, 2014 at Idaho National Laboratory, Idaho Falls, ID Outline: 1. Motivation 2. Tritium Plasma Experiment 3. INL/STAR's role on US-Japan collaboration 4. Role of TPE in PHENIX project 5. TPE modification and development of plasma-driven permeation M.Shimada | Tritium Focus

  4. Tritium 2016 11TH International Conference on Tritium Science...

    Office of Environmental Management (EM)

    Conference on Tritium Science and Technology More Documents & Publications Advanced Polymers for Tritium Service A New Hydrogen Processing Demonstration System Hydrogen Isotope...

  5. Tritium Focus Group Meeting

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

    Meeting Information Tritium Focus Group Charter (pdf) Hotel Information Classified Session Information Los Alamos Restaurants (pdf) LANL Information Visiting Los Alamos Area Map ...

  6. Tritium Plasma Experiment and

    Office of Environmental Management (EM)

    Plasma Experiment and its role in PHENIX program Masashi Shimada, Chase Taylor Fusion ... ID Outline: 1. Motivation 2. Tritium Plasma Experiment 3. INLSTAR's role on US-Japan ...

  7. Experiences with decontaminating tritium-handling apparatus

    SciTech Connect (OSTI)

    Maienschein, J.L.; Garcia, F.; Garza, R.G.; Kanna, R.L.; Mayhugh, S.R.; Taylor, D.T.

    1991-07-01

    Tritium-handling apparatus has been decontaminated as part of the shutdown of the LLNL Tritium Facility. Two stainless-steel gloveboxes that had been used to process lithium deuteride-tritide (LiDT) salt were decontaminated using the Portable Cleanup System so that they could be flushed with room air through the facility ventilation system. Further surface decontamination was performed by scrubbing the interior with paper towels and ethyl alcohol or Swish{trademark}. The surface contamination, as shown by swipe surveys, was reduced from 4{times}10{sup 4}--10{sup 6} disintegrations per minute (dpm)/cm{sup 2} to 2{times}10{sup 2}--4{times}10{sup 4} dpm/cm{sup 2}. Details on the decontamination operation are provided. A series of metal (palladium and vanadium) hydride storage beds have been drained of tritium and flushed with deuterium in order to remove as much tritium as possible. The bed draining and flushing procedure is described, and a calculational method is presented which allows estimation of the tritium remaining in a bed after it has been drained and flushed. Data on specific bed draining and flushing are given.

  8. Introduction to Airborne Tritium Tritides

    Office of Environmental Management (EM)

    ... Production industries for Tritium based products (Includes self-illuminating devices such as exit signs, gun sights, etc.) HOW DO TRITIUM TRITIDES ENTER THE ENVIRONMENT? Any item ...

  9. Methods for tritium labeling

    DOE Patents [OSTI]

    Andres, Hendrik; Morimoto, Hiromi; Williams, Philip G.

    1993-01-01

    Reagents and processes for reductively introducing deuterium or tritium into organic molecules are described. The reagents are deuterium or tritium analogs of trialkyl boranes, borane or alkali metal aluminum hydrides. The process involves forming these reagents in situ from alkali metal tritides or deuterides.

  10. Oxidative Tritium Decontamination System

    DOE Patents [OSTI]

    Gentile, Charles A. , Guttadora, Gregory L. , Parker, John J.

    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.

  11. Type A Accident Investigation Board report on the January 17, 1996, electrical accident with injury in Technical Area 21 Tritium Science and Fabrication Facility Los Alamos National Laboratory. Final report

    SciTech Connect (OSTI)

    1996-04-01

    An electrical accident was investigated in which a crafts person received serious injuries as a result of coming into contact with a 13.2 kilovolt (kV) electrical cable in the basement of Building 209 in Technical Area 21 (TA-21-209) in the Tritium Science and Fabrication Facility (TSFF) at Los Alamos National Laboratory (LANL). In conducting its investigation, the Accident Investigation Board used various analytical techniques, including events and causal factor analysis, barrier analysis, change analysis, fault tree analysis, materials analysis, and root cause analysis. The board inspected the accident site, reviewed events surrounding the accident, conducted extensive interviews and document reviews, and performed causation analyses to determine the factors that contributed to the accident, including any management system deficiencies. Relevant management systems and factors that could have contributed to the accident were evaluated in accordance with the guiding principles of safety management identified by the Secretary of Energy in an October 1994 letter to the Defense Nuclear Facilities Safety Board and subsequently to Congress.

  12. 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.

  13. Metabolism and dosimetry of tritium

    SciTech Connect (OSTI)

    Hill, R.L.; Johnson, J.R. )

    1993-12-01

    This document was prepared as a review of the current knowledge of tritium metabolism and dosimetry. The physical, chemical, and metabolic characteristics of various forms of tritium are presented as they pertain to performing dose assessments for occupational workers and for the general public. For occupational workers, the forms of tritium discussed include tritiated water, elemental tritium gas, skin absorption from elemental tritium gas-contaminated surfaces, organically bound tritium in pump oils, solvents and other organic compounds, metal tritides, and radioluminous paints. For the general public, age-dependent tritium metabolism is reviewed, as well as tritiated water, elemental tritium gas, organically bound tritium, organically bound tritium in food-stuffs, and tritiated methane. 106 refs.

  14. Tritium 2013 Presentation | Department of Energy

    Office of Environmental Management (EM)

    2013 Presentation More Documents & Publications Meeting Attendance - 32nd Tritium Focus Group Meeting Tritium Sessions At The 2012 ANS Meeting Overview of tritium activity in Japan...

  15. Tritium Instrument Demonstration Station (TIDS) | Department...

    Office of Environmental Management (EM)

    April 22-24, 2014. Tritium Instrument Demonstration Station (TIDS) (4.19 MB) More Documents & Publications Tritium Instrument Demonstration Station (TIDS) Tritium Instrument ...

  16. 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).

  17. Tritium waste package

    DOE Patents [OSTI]

    Rossmassler, Rich; Ciebiera, Lloyd; Tulipano, Francis J.; Vinson, Sylvester; Walters, R. Thomas

    1995-01-01

    A containment and waste package system for processing and shipping tritium xide waste received from a process gas includes an outer drum and an inner drum containing a disposable molecular sieve bed (DMSB) seated within 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 add oxygen present in the process gas to recombine to form water vapor, which is then adsorbed onto the DMSB.

  18. 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.

  19. PRODUCTION OF TRITIUM

    DOE Patents [OSTI]

    Jenks, G.H.; Shapiro, E.M.; Elliott, N.; Cannon, C.V.

    1963-02-26

    This invention relates to a process for the production of tritium by subjecting comminuted solid lithium fluoride containing the lithium isotope of atomic mass number 6 to neutron radiation in a self-sustaining neutronic reactor. The lithium fiuoride is heated to above 450 deg C. in an evacuated vacuum-tight container during radiation. Gaseous radiation products are withdrawn and passed through a palladium barrier to recover tritium. (AEC)

  20. Thin film tritium dosimetry

    DOE Patents [OSTI]

    Moran, Paul R.

    1976-01-01

    The present invention provides a method for tritium dosimetry. A dosimeter comprising a thin film of a material having relatively sensitive RITAC-RITAP dosimetry properties is exposed to radiation from tritium, and after the dosimeter has been removed from the source of the radiation, the low energy electron dose deposited in the thin film is determined by radiation-induced, thermally-activated polarization dosimetry techniques.

  1. Glovebox stripper system tritium capture efficiency-literature review

    SciTech Connect (OSTI)

    James, D. W.; Poore, A. S.

    2015-09-28

    Glovebox Stripper Systems (GBSS) are intended to minimize tritium emissions from glovebox confinement systems in Tritium facilities. A question was raised to determine if an assumed 99% stripping (decontamination) efficiency in the design of a GBBS was appropriate. A literature review showed the stated 99% tritium capture efficiency used for design of the GBSS is reasonable. Four scenarios were indicated for GBSSs. These include release with a single or dual stage setup which utilizes either single-pass or recirculation for stripping purposes. Examples of single-pass as well as recirculation stripper systems are presented and reviewed in this document.

  2. Tritium Permeation Activity at Safety and Tritium Applied

    Office of Environmental Management (EM)

    ... WS Reference: "Tritium Permeability of Incoloy 800H and Inconel 617" INLEXT-11-23265 and INLEXT-11-23265 rev1 Figure 12. Arrhenius plot of Incoloy 800H tritium permeability (FY ...

  3. Tritium Leak Detection: Strategies and Applications

    Office of Environmental Management (EM)

    Tritium Leak Detection: Strategies and Applications 2 *Radiological Protection *Elemental Tritium *Tritium Oxide *Locating the Point of Release *Real-time Measurement 3 The required radiological protection for tritium compounds is much higher than that for elemental tritium. 4 Elemental tritium will combine readily with oxygen to form tritium oxide and also will replace hydrogen atoms in compounds. 5 The presence of elemental tritium in the working environment is an indication of a

  4. 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.

  5. Facilities

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

    Facilities Facilities LANL's mission is to develop and apply science and technology to ensure the safety, security, and reliability of the U.S. nuclear deterrent; reduce global threats; and solve other emerging national security and energy challenges. Contact Operator Los Alamos National Laboratory (505) 667-5061 Some LANL facilities are available to researchers at other laboratories, universities, and industry. Unique facilities foster experimental science, support the Lab's security mission

  6. Facilities

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

    Secure and Sustainable Energy Future Mission/Facilities Facilities Tara Camacho-Lopez 2016-04-06T18:06:13+00:00 National Solar Thermal Test Facility (NSTTF) facility_nsttf_slide NSTTF's primary goal is to provide experimental engineering data for the design, construction, and operation of unique components and systems in proposed solar thermal electrical plants, which have three generic system architectures: line-focus (trough and continuous linear Fresnel reflector systems), point-focus central

  7. 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.

  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. Drum bubbler tritium processing system

    DOE Patents [OSTI]

    Rule, Keith; Gettelfinger, Geoff; Kivler, Paul

    1999-01-01

    A method of 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. 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.

  10. 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.

  11. Tritium Instrument Demonstration Station (TIDS)

    Broader source: Energy.gov [DOE]

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

  12. 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.

  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. Monitoring of tritium

    DOE Patents [OSTI]

    Corbett, James A.; Meacham, Sterling A.

    1981-01-01

    The fluid from a breeder nuclear reactor, which may be the sodium cooling fluid or the helium reactor-cover-gas, or the helium coolant of a gas-cooled reactor passes over the portion of the enclosure of a gaseous discharge device which is permeable to hydrogen and its isotopes. The tritium diffused into the discharge device is radioactive producing beta rays which ionize the gas (argon) in the discharge device. The tritium is monitored by measuring the ionization current produced when the sodium phase and the gas phase of the hydrogen isotopes within the enclosure are in equilibrium.

  15. 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

  16. 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.

  17. Recent progress on tritium technology research and development for a fusion reactor in Japan Atomic Energy Agency

    SciTech Connect (OSTI)

    Hayashi, T.; Nakamura, H.; Kawamura, Y.; Iwai, Y.; Isobe, K.; Yamada, M.; Kurata, R.; Edao, Y.; Suzuki, T.; Oyaizu, M.; Yamanishi, T.

    2015-03-15

    JAEA (Japan Atomic Energy Agency) manages 2 tritium handling laboratories: Tritium Processing Laboratory (TPL) in Tokai and DEMO-RD building in Rokkasho. TPL has been accumulating a gram level tritium safety handling experiences without any accidental tritium release to the environment for more than 25 years. Recently, our activities have focused on 3 categories, as follows. First, the development of a detritiation system for ITER. This task is the demonstration test of a wet Scrubber Column (SC) as a pilot scale (a few hundreds m{sup 3}/h of processing capacity). Secondly, DEMO-RD tasks are focused on investigating the general issues required for DEMO-RD design, such as structural materials like RAFM (Reduced Activity Ferritic/Martensitic steels) and SiC/SiC, functional materials like tritium breeder and neutron multiplier, and tritium. For the last 4 years, we have spent a lot of time and means to the construction of the DEMO-RD facility and to its licensing, so we have just started the actual research program with tritium and other radioisotopes. This tritium task includes tritium accountancy, tritium basic safety research such as tritium interactions with various materials, which will be used for DEMO-RD and durability. The third category is the recovery work from the Great East Japan earthquake (2011 earthquake). It is worth noting that despite the high magnitude of the earthquake, TPL was able to confine tritium properly without any accidental tritium release.

  18. 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

  19. Facilities

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

    Facilities The the WTGa1 turbine (aka DOE/SNL #1) retuns to power as part of a final series of commissioning tests. Permalink Gallery First Power for SWiFT Turbine Achieved during Recommissioning Facilities, News, Renewable Energy, SWIFT, Wind Energy, Wind News First Power for SWiFT Turbine Achieved during Recommissioning The Department of Energy's Scaled Wind Farm Technology (SWiFT) Facility reached an exciting milestone with the return to power production of the WTGa1 turbine (aka DOE/SNL #1)

  20. Tritium on Metal Surfaces | Department of Energy

    Office of Environmental Management (EM)

    on Metal Surfaces Tritium on Metal Surfaces Presentation from the 34th Tritium Focus Group Meeting held in Idaho Falls, Idaho on September 23-25, 2014. Tritium on Metal Surfaces (1.03 MB) More Documents & Publications Modeling Tritium on Metal Surfaces Tritium Plasma Experiment and Its Role in PHENIX Program Light Water Detritiation using the CECE Process

  1. Fermilab | Tritium at Fermilab | Tritium in Surface Water

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

    Surface Water Fermilab map Fermilab has conducted an environmental monitoring program on site for roughly 40 years. In November of 2005, for the first time, we detected low levels of tritium in Indian Creek, one of three creeks that travel through the Fermilab site. Low but measurable levels of tritium continue to be detected in Indian Creek. All tritium levels found on site are well below any federal health and environmental standards. The Department of Energy standard for surface water is

  2. Dismantling of the PETRA glove box: tritium contamination and inventory assessment

    SciTech Connect (OSTI)

    Wagner, R.

    2015-03-15

    The PETRA facility is the first installation in which experiments with tritium were carried out at the Tritium Laboratory Karlsruhe. After completion of two main experimental programs, the decommissioning of PETRA was initiated with the aim to reuse the glove box and its main still valuable components. A decommissioning plan was engaged to: -) identify the source of tritium release in the glove box, -) clarify the status of the main components, -) assess residual tritium inventories, and -) de-tritiate the components to be disposed of as waste. Several analytical techniques - calorimetry on small solid samples, wipe test followed by liquid scintillation counting for surface contamination assessment, gas chromatography on gaseous samples - were deployed and cross-checked to assess the remaining tritium inventories and initiate the decommissioning process. The methodology and the main outcomes of the numerous different tritium measurements are presented and discussed. (authors)

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

    Office of Environmental Management (EM)

    Laboratory (12.9 MB) More Documents & Publications Tritium Plasma Experiment and Its Role in PHENIX Program Technological Assessment of Plasma Facing Components for DEMO Reactors

  4. 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.

  5. Particulate Generation in Tritium Systems

    Office of Environmental Management (EM)

    Particulate Generation in a Tritium System Paul Cloessner, PhD Laboratory Fellow Tritium Focus Group February 22, 2014 Outline * Description of Events * Analysis of Material * Sources of material contamination * System Restoration/Modifications * Contaminant Minimization and Control * Lessons Learned 2 An Unpleasant Surprise * Let down filter on compressor became plugged after 10 years of operation. * Tritium processing interrupted when other filters (flow orifices) became plugged approximately

  6. 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.

  7. Tritium Handling and Safe Storage

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

    ... Individual mm Millimeter mrem Millirem NFPA National Fire Protection Association NP ... Handling of Tritium, published in 1991; and U.S. Department of Energy (DOE) publications. ...

  8. Tritium Handling and Safe Storage

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

    ... Level mm Millimeter mrem Millirem NFPA National Fire Protection Association NMMSS ... Safe Handling of Tritium, published in 1991, in addition to the French Nuclear Safety ...

  9. Tritium Handling and Safe Storage

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

    ... Individual mm Millimeters mrem Millirem NFPA National Fire Protection Association NPDWR ... "Safe Handling of Tritium," published in 1991; and U.S. Department of Energy (DOE) ...

  10. 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 ...

  11. The certification process for tritium operators at TFTR

    SciTech Connect (OSTI)

    Gentile, C.A.; Murphy, S.E.; LaMarche, P.H.; Contino, A.M.; Gordon, S.

    1995-12-31

    The TFTR project, in concert with the PPPL Office of Certification and Training (OC and T), has established a program by which Tritium Operations Personnel are certified for their respective positions in accordance with DOE Order 5480.20A Personnel Selection, Qualification, and Training at DOE Nuclear Facilities and DOE Order 5480.19 conduct of Operations Requirements for DOE Facilities. The certification process commences during the candidate`s interview for the position of TFTR Tritium Operator. Prior to accepting a candidate into the tritium operation program, a detailed educational and work experience record is constructed for the candidate, including an interview by OC and T personnel to assess the candidates credentials and ability to successfully complete the program. The typical successful candidate for the position of TFTR Tritium Operator has worked in the nuclear or chemical industry for several years, and in many cases possess a college degree. A US Nuclear Navy background is quite common for many of the applicants. Candidates complete the program in 4 to 6 months, and typically move into supervisory positions (Tritium Shift Supervisors) within 2 to 3 years.

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

    Office of Environmental Management (EM)

    HFBR Tritium Brookhaven National Laboratory - HFBR Tritium January 1, 2014 - 12:00pm ... InstallationName, State: Brookhaven National Laboratory Responsible DOE Office: Office of ...

  13. 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.

  14. Tritium accountancy in fusion systems

    SciTech Connect (OSTI)

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

    2015-03-15

    The US Department of Energy (DOE) has clearly defined requirements for nuclear material control and accountability (MCA) of tritium whereas the International Atomic Energy Agency (IAEA) does not since tritium is not a fissile material. MCA 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 (MBA) are defined to aid in the tracking and reporting of nuclear material movements and inventories. Material sub-accounts (MSA) are established along with key measurement points (KMP) 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 MSA. 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 breeding, 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. (authors)

  15. 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.

  16. 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.

  17. Low tritium partial pressure permeation system for mass transport measurement in lead lithium eutectic

    SciTech Connect (OSTI)

    Pawelko, R. J.; Shimada, M.; Katayama, K.; Fukada, S.; Humrickhouse, P. W.; Terai, T.

    2015-11-28

    This paper describes a new experimental system designed to investigate tritium mass transfer properties in materials important to fusion technology. Experimental activities were carried out at the Safety and Tritium Applied Research (STAR) facility located at the Idaho National Laboratory (INL). The tritium permeation measurement system was developed as part of the Japan/US TITAN collaboration to investigate tritium mass transfer properties in liquid lead lithium eutectic (LLE) alloy. The experimental system is configured to measure tritium mass transfer properties at low tritium partial pressures. Initial tritium permeation scoping tests were conducted on a 1 mm thick α-Fe plate to determine operating parameters and to validate the experimental technique. A second series of permeation tests was then conducted with the α-Fe plate covered with an approximately 8.5 mm layer of liquid lead lithium eutectic alloy (α-Fe/LLE). We present preliminary tritium permeation data for α-Fe and α-Fe/LLE at temperatures between 400 and 600°C and at tritium partial pressures between 1.7E-3 and 2.5 Pa in helium. Preliminary results for the α-Fe plate and α-Fe/LLE indicate that the data spans a transition region between the diffusion-limited regime and the surface-limited regime. In conclusion, additional data is required to determine the existence and range of a surface-limited regime.

  18. Low tritium partial pressure permeation system for mass transport measurement in lead lithium eutectic

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

    Pawelko, R. J.; Shimada, M.; Katayama, K.; Fukada, S.; Humrickhouse, P. W.; Terai, T.

    2015-11-28

    This paper describes a new experimental system designed to investigate tritium mass transfer properties in materials important to fusion technology. Experimental activities were carried out at the Safety and Tritium Applied Research (STAR) facility located at the Idaho National Laboratory (INL). The tritium permeation measurement system was developed as part of the Japan/US TITAN collaboration to investigate tritium mass transfer properties in liquid lead lithium eutectic (LLE) alloy. The experimental system is configured to measure tritium mass transfer properties at low tritium partial pressures. Initial tritium permeation scoping tests were conducted on a 1 mm thick α-Fe plate to determinemore » operating parameters and to validate the experimental technique. A second series of permeation tests was then conducted with the α-Fe plate covered with an approximately 8.5 mm layer of liquid lead lithium eutectic alloy (α-Fe/LLE). We present preliminary tritium permeation data for α-Fe and α-Fe/LLE at temperatures between 400 and 600°C and at tritium partial pressures between 1.7E-3 and 2.5 Pa in helium. Preliminary results for the α-Fe plate and α-Fe/LLE indicate that the data spans a transition region between the diffusion-limited regime and the surface-limited regime. In conclusion, additional data is required to determine the existence and range of a surface-limited regime.« less

  19. 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.

  20. 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

  1. Fermilab | Tritium at Fermilab | Tritium in Sanitary Sewers

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

    This level is below the Department of Energy standard for tritium in sanitary sewers, which is 9,500 pCiml. The lab also has to meet a new DOE standard of five curies (or five ...

  2. Differential atmospheric tritium sampler

    DOE Patents [OSTI]

    Griesbach, Otto A.; Stencel, Joseph R.

    1990-01-01

    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 mixture 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.

  3. 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.

  4. WASTE CHARACTERIZATION OF POLYMERIC COMPONENTS EXPOSED TO TRITIUM GAS

    SciTech Connect (OSTI)

    Clark, E

    2008-02-15

    A recent independent review led to uncertainty about the technical basis for characterizing the residual amount of tritium in polymer components used in the Savannah River Site Tritium Facilities that are sent for waste disposal. A review of a paper published in the open literature firmly establishes the basis of the currently used characterization, 10 Ci/cc. Information provided in that paper about exposure experiments performed at the DOE Mound Laboratory allows the calculation of the currently used characterization. These experiments involved exposure of high density polyethylene (HD-PE) to initially 1 atm tritium gas. In addition, a review of recent research at the Savannah River Site not only further substantiates this characterization, but also establishes its use for ultra-high molecular weight polyethylene (UHMW-PE), polytetrafluoroethylene (PTFE, a trade name is Teflon{reg_sign}), and Vespel{reg_sign} polyimide. 10 Ci/cc tritium is a representative characterization for any type of polymer components exposed at ambient temperature and at approximately 1 atm. tritium gas.

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

    Office of Environmental Management (EM)

    Fusion Nuclear Science and Technology Program - Status and Plans for Tritium Research Tritium Plasma Experiment and Its Role in PHENIX Program Tritium research activities in Safety ...

  6. Normalized Tritium Quantification Approach (NoTQA) a Method for Quantifying Tritium Contaminated Trash and Debris at LLNL

    SciTech Connect (OSTI)

    Dominick, J L; Rasmussen, C L

    2008-07-23

    Several facilities and many projects at LLNL work exclusively with tritium. These operations have the potential to generate large quantities of Low-Level Radioactive Waste (LLW) with the same or similar radiological characteristics. A standardized documented approach to characterizing these waste materials for disposal as radioactive waste will enhance the ability of the Laboratory to manage them in an efficient and timely manner while ensuring compliance with all applicable regulatory requirements. This standardized characterization approach couples documented process knowledge with analytical verification and is very conservative, overestimating the radioactivity concentration of the waste. The characterization approach documented here is the Normalized Tritium Quantification Approach (NoTQA). This document will serve as a Technical Basis Document which can be referenced in radioactive waste characterization documentation packages such as the Information Gathering Document. In general, radiological characterization of waste consists of both developing an isotopic breakdown (distribution) of radionuclides contaminating the waste and using an appropriate method to quantify the radionuclides in the waste. Characterization approaches require varying degrees of rigor depending upon the radionuclides contaminating the waste and the concentration of the radionuclide contaminants as related to regulatory thresholds. Generally, as activity levels in the waste approach a regulatory or disposal facility threshold the degree of required precision and accuracy, and therefore the level of rigor, increases. In the case of tritium, thresholds of concern for control, contamination, transportation, and waste acceptance are relatively high. Due to the benign nature of tritium and the resulting higher regulatory thresholds, this less rigorous yet conservative characterization approach is appropriate. The scope of this document is to define an appropriate and acceptable

  7. Combined gettering and molten salt process for tritium recovery from lithium

    SciTech Connect (OSTI)

    Sze, D.K.; Finn, P.A.; Bartlit, J.; Tanaka, S.; Teria, T.; Yamawaki, M.

    1988-02-01

    A new tritium recovery concept from lithium has been developed as part of the US/Japan collaboration on Reversed-Field Pinch Reactor Design Studies. This concept combines the ..gamma..-gettering process as the front end to recover tritium from the coolant, and a molten salt recovery process to extract tritium for fuel processing. A secondary lithium is used to regenerate the tritium from the gettering bed and, in the process, increases the tritium concentration by a factor of about 20. That way, the required size of the molten salt process becomes very small. A potential problem is the possible poisoning of the gettering bed by the salt dissolved in lithium. 16 refs., 6 figs.

  8. A study of tritium in municipal solid waste leachate and gas

    SciTech Connect (OSTI)

    Mutch Jr, R. D.; Mahony, J. D.

    2008-07-15

    It has become increasingly clear in the last few years that the vast majority of municipal solid waste landfills produce leachate that contains elevated levels of tritium. The authors recently conducted a study of landfills in New York and New Jersey and found that the mean concentration of tritium in the leachate from ten municipal solid waste (MSW) landfills was 33,800 pCi/L with a peak value of 192,000 pCi/L. A 2003 study in California reported a mean tritium concentration of 99,000 pCi/L with a peak value of 304,000 pCi/L. Studies in Pennsylvania and the UK produced similar results. The USEPA MCL for tritium is 20,000 pCi/L. Tritium is also manifesting itself as landfill gas and landfill gas condensate. Landfill gas condensate samples from landfills in the UK and California were found to have tritium concentrations as high as 54,400 and 513,000 pCi/L, respectively. The tritium found in MSW leachate is believed to derive principally from gaseous tritium lighting devices used in some emergency exit signs, compasses, watches, and even novelty items, such as 'glow stick' key chains. This study reports the findings of recent surveys of leachate from a number of municipal solid waste landfills, both open and closed, from throughout the United States and Europe. The study evaluates the human health and ecological risks posed by elevated tritium levels in municipal solid waste leachate and landfill gas and the implications to their safe management. We also assess the potential risks posed to solid waste management facility workers exposed to tritium-containing waste materials in transfer stations and other solid waste management facilities. (authors)

  9. Separation of Tritium from Wastewater

    SciTech Connect (OSTI)

    JEPPSON, D.W.

    2000-01-25

    A proprietary tritium loading bed developed by Molecular Separations, Inc (MSI) has been shown to selectively load tritiated water as waters of hydration at near ambient temperatures. Tests conducted with a 126 {micro}C{sub 1} tritium/liter water standard mixture showed reductions to 25 {micro}C{sub 1}/L utilizing two, 2-meter long columns in series. Demonstration tests with Hanford Site wastewater samples indicate an approximate tritium concentration reduction from 0.3 {micro}C{sub 1}/L to 0.07 {micro}C{sub 1}/L for a series of two, 2-meter long stationary column beds Further reduction to less than 0.02 {micro}C{sub 1}/L, the current drinking water maximum contaminant level (MCL), is projected with additional bed media in series. Tritium can be removed from the loaded beds with a modest temperature increase and the beds can be reused Results of initial tests are presented and a moving bed process for treating large quantities of wastewaters is proposed. The moving bed separation process appears promising to treat existing large quantities of wastewater at various US Department of Energy (DOE) sites. The enriched tritium stream can be grouted for waste disposition. The separations system has also been shown to reduce tritium concentrations in nuclear reactor cooling water to levels that allow reuse. Energy requirements to reconstitute the loading beds and waste disposal costs for this process appear modest.

  10. Production of highly tritiated water for tritium exposure studies

    SciTech Connect (OSTI)

    Muirhead, C.; Pilatzke, K.; Tripple, A.; Philippi, N.; McCrimmon, K.; Castillo, I.; Boniface, H.; Suppiah, S.

    2015-03-15

    Tritium Facility staff at Chalk River Laboratories (CRL) have successfully prepared highly tritiated water for use in radiation resistance of PEM (Proton Exchange Membrane-based)electrolyser membrane. The goal of System A was to convert a known amount of elemental tritium (HT) into tritiated water vapour using a copper(II) oxide bed, and to condense the tritiated water vapour into a known amount of chilled heavy water (D{sub 2}O). The conversion and capture of tritium using this system is close to 100%. The goal of System B was to transfer tritiated water from the containment vessel to an exposure vessel (experiment) in a controlled and safe manner. System B is based on the pushing of D{sub 2}0 with low-pressure argon carrier gas to a calibrated volume and then to the exposure vessel. A method for delivering a known and controlled amount of tritiated water has been successfully demonstrated at CRL. Using both systems Tritium Facility staff have made and distributed highly tritiated water in a safe and controlled manner. This paper focuses on how the tritiated water was produced and dispensed to the experiment.

  11. Primer on tritium safe handling practices

    SciTech Connect (OSTI)

    Not Available

    1994-12-01

    This Primer is designed for use by operations and maintenance personnel to improve their knowledge of tritium safe handling practices. It is applicable to many job classifications and can be used as a reference for classroom work or for self-study. It is presented in general terms for use throughout the DOE Complex. After reading it, one should be able to: describe methods of measuring airborne tritium concentration; list types of protective clothing effective against tritium uptake from surface and airborne contamination; name two methods of reducing the body dose after a tritium uptake; describe the most common method for determining amount of tritium uptake in the body; describe steps to take following an accidental release of airborne tritium; describe the damage to metals that results from absorption of tritium; explain how washing hands or showering in cold water helps reduce tritium uptake; and describe how tritium exchanges with normal hydrogen in water and hydrocarbons.

  12. Mercury and tritium removal from DOE waste oils

    SciTech Connect (OSTI)

    Klasson, E.T.

    1997-10-01

    This work covers the investigation of vacuum extraction as a means to remove tritiated contamination as well as the removal via sorption of dissolved mercury from contaminated oils. The radiation damage in oils from tritium causes production of hydrogen, methane, and low-molecular-weight hydrocarbons. When tritium gas is present in the oil, the tritium atom is incorporated into the formed hydrocarbons. The transformer industry measures gas content/composition of transformer oils as a diagnostic tool for the transformers` condition. The analytical approach (ASTM D3612-90) used for these measurements is vacuum extraction of all gases (H{sub 2}, N{sub 2}, O{sub 2}, CO, CO{sub 2}, etc.) followed by analysis of the evolved gas mixture. This extraction method will be adapted to remove dissolved gases (including tritium) from the SRS vacuum pump oil. It may be necessary to heat (60{degrees}C to 70{degrees}C) the oil during vacuum extraction to remove tritiated water. A method described in the procedures is a stripper column extraction, in which a carrier gas (argon) is used to remove dissolved gases from oil that is dispersed on high surface area beads. This method appears promising for scale-up as a treatment process, and a modified process is also being used as a dewatering technique by SD Myers, Inc. (a transformer consulting company) for transformers in the field by a mobile unit. Although some mercury may be removed during the vacuum extraction, the most common technique for removing mercury from oil is by using sulfur-impregnated activated carbon (SIAC). SIAC is currently being used by the petroleum industry to remove mercury from hydrocarbon mixtures, but the sorbent has not been previously tested on DOE vacuum oil waste. It is anticipated that a final process will be similar to technologies used by the petroleum industry and is comparable to ion exchange operations in large column-type reactors.

  13. Laser-assisted isotope separation of tritium

    DOE Patents [OSTI]

    Herman, Irving P. (Castro Valley, CA); Marling, Jack B. (Livermore, CA)

    1983-01-01

    Methods for laser-assisted isotope separation of tritium, using infrared multiple photon dissociation of tritium-bearing products in the gas phase. One such process involves the steps of (1) catalytic exchange of a deuterium-bearing molecule XYD with tritiated water DTO from sources such as a heavy water fission reactor, to produce the tritium-bearing working molecules XYT and (2) photoselective dissociation of XYT to form a tritium-rich product. By an analogous procedure, tritium is separated from tritium-bearing materials that contain predominately hydrogen such as a light water coolant from fission or fusion reactors.

  14. 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.

  15. Facility Floorplan

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

    facility floorplan Facility Floorplan

  16. Modeling Tritium on Metal Surfaces | Department of Energy

    Office of Environmental Management (EM)

    Tritium on Metal Surfaces Modeling Tritium on Metal Surfaces Presentation from the 36th Tritium Focus Group Meeting held in Los Alamos, New Mexico, November 3-5, 2015. Modeling Tritium on Metal Surfaces (6.14 MB) More Documents & Publications Tritium on Metal Surfaces DOE-HDBK-1079-94 Overview of tritium activity in Japan

  17. 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}.

  18. 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.

  19. 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.

  20. A Plan for Modularization of Tritium Components

    Office of Environmental Management (EM)

    Tritium Components Randy Davis Davis Consultants M-TRT-H-00089 Savannah River Nuclear Solutions, LLC April 22, 2014 M-TRT-H-00089 Current Approach * All "tritium wetted components" ...

  1. Fermilab | Tritium at Fermilab | Frequently asked questions

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

    To date, the levels of tritium Fermilab has detected in Indian Creek at the site boundary have not risen above single digits. The DOE standard for tritium in sanitary sewer water ...

  2. 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.

  3. 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.

  4. Fermilab | Tritium at Fermilab | Kress Creek Results

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

    Kress Creek Results chart This chart (click chart for larger version) shows the levels of tritium in Kress Creek since January 2006. To date, Fermilab has not detected tritium in Kress Creek. The detection limit is one picocurie per milliliter (see footnote). Increased monitoring began on Kress Creek following detection of low levels of tritium in Indian Creek in November 2005. The levels of tritium measured in the Fermilab cooling ponds and in Indian Creek are well below federal water standards

  5. Tritium Gas Processing for Magnetic Fusion

    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.

  6. Bulk Tritium Shipping Package Overview and Status

    Broader source: Energy.gov [DOE]

    Presentation from the 36th Tritium Focus Group Meeting held in Los Alamos, New Mexico, November 3-5, 2015.

  7. 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.

  8. A Plan for Modularization of Tritium Components

    Broader source: Energy.gov [DOE]

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

  9. 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.

  10. Design of the Target Fabrication Tritium Laboratory

    SciTech Connect (OSTI)

    Sherohman, J.W.; Roberts, D.H.; Levine, B.H.

    1982-05-05

    The design of the Target Fabrication Tritium Laboratory for deuterium-tritium fuel processing for laser fusion targets has been accomplished with the intent of providing redundant safeguard systems. The design of the tritium laboratory is based on a combination of tritium handling techniques that are currently used by experienced laboratories. A description of the laboratory in terms of its interrelated processing systems is presented to provide an understanding of the design features for safe operation.

  11. Tritium contamination and decontamination of sealing oil for vacuum pump

    SciTech Connect (OSTI)

    Takeishi, T.; Kotoh, K.; Kawabata, Y.; Tanaka, J.I.; Kawamura, S.; Iwata, M.

    2015-03-15

    The existence of tritium-contaminated oils from vacuum pumps used in tritium facilities, is becoming an important issue since there is no disposal way for tritiated waste oils. On recovery of tritiated water vapor in gas streams, it is well-known that the isotope exchange reaction between the gas phase and the liquid phase occurs effectively at room temperature. We have carried out experiments using bubbles to examine the tritium contamination and decontamination of a volume of rotary-vacuum-pump oil. The contamination of the pump oil was made by bubbling tritiated water vapor and tritiated hydrogen gas into the oil. Subsequently the decontamination was processed by bubbling pure water vapor and dry argon gas into the tritiated oil. Results show that the water vapor bubbling was more effective than dry argon gas. The experiment also shows that the water vapor bubbling in an oil bottle can remove and transfer tritium efficiently from the tritiated oil into another water-bubbling bottle.

  12. Characterization of LaNi{sub 4.25}Al{sub 0.75} tritide for use as a long term tritium storage medium

    SciTech Connect (OSTI)

    Wermer, J.R.

    1994-10-01

    Applications of metal hydride technology has offered numerous safety as well as operating advantages for tritium processing operations. The SRS Replacement Tritium Facility (RTF) utilizes this technology extensively. During design of the RTF systems, LaNi{sub 4.25}Al{sub 0.75} (LANA.75) was chosen as the primary tritium storage material. This material was selected largely because of the isotherm plateau pressure, which allows the tritium to be stored as a metal tritide at subatmospheric pressures while still being able to generate pressures of >1000 mm Hg needed for process applications. A benefit of this substitution is an increase in the stability of this material to tritium aging effects and to disproportionation. The LANA.75 material, like many metal tritides used for tritium processing, retains insoluble helium-3 which is born in the metal lattice through radiolytic decay of tritium. This causes changes in the thermodynamics of the metal-hydrogen system, decreasing the {alpha}-{beta} plateau pressure, increasing the plateau slope, and decreasing the reversible hydriding capacity. The latter also includes the growth of a tritium {open_quotes}heel{close_quotes} which cannot be removed under normal processing conditions. All of these factors affect the long-term performance of LANA.75-tritide in processing applications. Tritium aging studies have been underway on LANA.75 since 1987 in the SRTC Materials Test Facility. Material characterization of LANA.75-tritide has been completed on material exposed to tritium for 5.4 years at full stoichiometry.

  13. 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.

  14. EVALUATION OF ALTERNATE STAINLESS STEEL SURFACE TREATMENTS FOR MASS SPECTROSCOPY AND OTHER TRITIUM SYSTEMS

    SciTech Connect (OSTI)

    Clark, E.; Mauldin, C.; Neikirk, K.

    2012-02-29

    There are specific components in the SRS Tritium Facilities that are required to introduce as few chemical impurities (such as protium and methane) as possible into the process gas. Two such components are the inlet systems for the mass spectroscopy facilities and hydrogen isotope mix standard containers. Two vendors now passivate stainless steel components for these systems, and both are relatively small businesses whose future viability can be questioned, which creates the need for new sources. Stainless steel containers were designed to evaluate alternate surface treatment vendors for tritium storage and handling for these high purity tritium systems. Five vendors applied their own 'best' surface treatments to two containers each - one was a current vendor, another was a chemical vapor deposited silicon coating, and the other three were electropolishing and chemical cleaning vendors. Pure tritium gas was introduced into all ten containers and the composition was monitored over time. The only observed impurities in the gas were some HT, less CT{sub 4}, and very small amounts of T{sub 2}O in all cases. The currently used vendor treated containers contained the least impurities. The chemical vapor deposited silicon treatment resulted in the highest impurity levels. Sampling one set of containers after about one month of tritium exposure revealed the impurity level to be nearly the same as that after more than a year of exposure - this result suggests that cleaning new stainless steel components by tritium gas contact for about a month may be a worthy operation.

  15. Tritium Waste Treatment System component failure data analysis from June 18, 1984--December 31, 1989

    SciTech Connect (OSTI)

    Cadwallader, L.C. ); Stolpe Gavett, M.A. )

    1990-09-01

    This document gives the failure rates for the major tritium-bearing components in the Tritium Waste Treatment System at the Tritium Systems Test Assembly, which is a fusion research and technology facility at the Los Alamos National Laboratory. The failure reports, component populations, and operating demands/hours are given in this report, and sample calculations for binomial demand failure rates and poisson hourly failure rates are given in the appendices. The failure rates for tritium-bearing components were on the order of the screening failure rate values suggested for fusion reliability and risk analyses. More effort should be directed toward collecting and analyzing fusion component failure data, since accurate failure rates are necessary to refine reliability and risk analyses. 15 refs., 4 figs., 4 tabs.

  16. Meeting Attendance - 32nd Tritium Focus Group Meeting | Department of

    Office of Environmental Management (EM)

    Energy 2nd Tritium Focus Group Meeting Meeting Attendance - 32nd Tritium Focus Group Meeting Attendees to the 32nd Tritium Focus Group Meeting held in Germantown, Maryland, April 23-25, 2013. Meeting Attendance - 32nd Tritium Focus Group Meeting (76.56 KB) More Documents & Publications Meeting Attendance - 33rd Tritium Focus Group Meeting Meeting Attendance - 34th Tritium Focus Group Meeting Tritium 2013 Presentation

  17. 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

  18. EIS-0271: Notice of Intent To Prepare an Environmental Impact...

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

    Construction and Operation of a Tritium Extraction Facility at the Savannah River Site The ... construction and operation of a Tritium Extraction Facility (TEF) pursuant to the National ...

  19. Sorption of tritium and tritiated water on construction materials

    SciTech Connect (OSTI)

    Dickson, R.S.; Miller, J.M. . Chalk River Nuclear Labs.)

    1992-03-01

    In this paper, sorption and desorption of tritium (HT) and tritiated water (HTO) on materials to be used in the construction of fusion facilities are studied. In ca. 24-hour exposures in argon or room air, metal samples sorbed 8-200 {mu}Ci/m{sup 2} (1 Ci = 3.7 {times} 10{sup 10} Bq) of tritium form atmospheres of 5-9 Ci/m{sup 3} HT, and non-metallic samples sorbed 60-800 {mu}Ci/m{sup 2} from atmospheres of 14 Ci/m{sup 3} HT. Sorption of HTO varied much more widely than HT sorption for different samples, ranging from 4 {mu}Ci/m{sup 2} for glass to 1,300,000 {mu}Ci/m{sup 3} HTO in room air. Time dependence of desorption in dry air showed a rapid initial process and a slower secondary process.

  20. Titanium for long-term tritium storage

    SciTech Connect (OSTI)

    Heung, L.K.

    1994-12-01

    Due to the reduction of nuclear weapon stockpile, there will be an excess of tritium returned from the field. The excess tritium needs to be stored for future use, which might be several years away. A safe and cost effective means for long term storage of tritium is needed. Storing tritium in a solid metal tritide is preferred to storing tritium as a gas, because a metal tritide can store tritium in a compact form and the stored tritium will not be released until heat is applied to increase its temperature to several hundred degrees centigrade. Storing tritium as a tritide is safer and more cost effective than as a gas. Several candidate metal hydride materials have been evaluated for long term tritium storage. They include uranium, La-Ni-Al alloys, zirconium and titanium. The criteria used include material cost, radioactivity, stability to air, storage capacity, storage pressure, loading and unloading conditions, and helium retention. Titanium has the best combination of properties and is recommended for long term tritium storage.

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

    SciTech Connect (OSTI)

    Clark, E; Kirk Shanahan, K

    2006-05-31

    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

  2. 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.

  3. 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.

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

    Broader source: Energy.gov [DOE]

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

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

    Broader source: Energy.gov [DOE]

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

  6. 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.

  7. CHARTER OF THE TRITIUM FOCUS GROUP (TFG)

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

    CHARTER 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, handling, transportation, storage, and operations, and to enhance communication across the Department of Energy (DOE) (inclusive of the National Nuclear Security Administration (NNSA)) on all matters related to tritium. OBJECTIVES - The objectives of the TFG include: 1. Serving as an efficient forum for communication and

  8. TRITIUM IN-BED ACCOUNTABILITY FOR A PASSIVELY COOLED, ELECTRICALLY HEATED HYDRIDE BED

    SciTech Connect (OSTI)

    Klein, J.; Foster, P.

    2011-01-21

    A PAssively Cooled, Electrically heated hydride (PACE) Bed has been deployed into tritium service in the Savannah River Site (SRS) Tritium Facilities. The bed design, absorption and desorption performance, and cold (non-radioactive) in-bed accountability (IBA) results have been reported previously. Six PACE Beds were fitted with instrumentation to perform the steady-state, flowing gas calorimetric inventory method. An IBA inventory calibration curve, flowing gas temperature rise ({Delta}T) versus simulated or actual tritium loading, was generated for each bed. Results for non-radioactive ('cold') tests using the internal electric heaters and tritium calibration results are presented. Changes in vacuum jacket pressure significantly impact measured IBA {Delta}T values. Higher jacket pressures produce lower IBA {Delta}T values which underestimate bed tritium inventories. The exhaust pressure of the IBA gas flow through the bed's U-tube has little influence on measured IBA {Delta}T values, but larger gas flows reduce the time to reach steady-state conditions and produce smaller tritium measurement uncertainties.

  9. Tritium High Vacuum Pump Test Plan

    Office of Environmental Management (EM)

    High Vacuum Pump Test Plan Tritium Programs Engineering Louis Boone Joel Bennett ... Shimming will have to be internal to the pump. Test System Measure ultimate vacuum with ...

  10. CHARTER OF THE TRITIUM FOCUS GROUP (TFG)

    Office of Environmental Management (EM)

    and coordination of tritium issues and activities across the DOE complex and beyond. 2. Promoting sharing and application of state-of-the-art design and engineering...

  11. 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.

  12. Tritium High Vacuum Pump Test Plan | Department of Energy

    Office of Environmental Management (EM)

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

  13. Meeting Attendance - 35th Tritium Focus Group Meeting | Department of

    Office of Environmental Management (EM)

    Energy 5th Tritium Focus Group Meeting Meeting Attendance - 35th Tritium Focus Group Meeting Attendees to the 35th Tritium Focus Group Meeting held in Princeton, NJ on May 5-7, 2015. Meeting Attendance - 35th Tritium Focus Group Meeting (108.18 KB) More Documents & Publications Meeting Attendance - 34th Tritium Focus Group Meeting Meeting Attendance - 33rd Tritium Focus Group Meeting

  14. First Irradiated Tritium Rods Arrive At SRS | National Nuclear...

    National Nuclear Security Administration (NNSA)

    Tritium is a radioactive isotope of the element hydrogen. Tritium decays at about five percent per year and therefore must be periodically replaced in nuclear weapons. ...

  15. SILC target design for accelerator production of tritium (APT...

    Office of Scientific and Technical Information (OSTI)

    SILC target design for accelerator production of tritium (APT) Citation Details In-Document Search Title: SILC target design for accelerator production of tritium (APT) An ...

  16. 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...

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

    Office of Environmental Management (EM)

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

  18. Luis Alvarez, the Hydrogen Bubble Chamber, Tritium, and Dinosaurs

    Office of Scientific and Technical Information (OSTI)

    Luis Alvarez, the Hydrogen Bubble Chamber, Tritium, and Dinosaurs Resources with ... standard of length, co-discovered the hydrogen isotope tritium, searched for hidden ...

  19. In-Reactor Measurement of Tritium Permeation through Stainless...

    Office of Environmental Management (EM)

    In-Reactor Measurement of Tritium Permeation through Stainless Steel Cladding In-Reactor Measurement of Tritium Permeation through Stainless Steel Cladding Presentation from the ...

  20. 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...

  1. TRITIUM EFFECTS ON WELDMENT FRACTURE TOUGHNESS

    SciTech Connect (OSTI)

    Morgan, M; Michael Tosten, M; Scott West, S

    2006-07-17

    The effects of tritium on the fracture toughness properties of Type 304L stainless steel and its weldments were measured. Fracture toughness data are needed for assessing tritium reservoir structural integrity. This report provides data from J-Integral fracture toughness tests on unexposed and tritium-exposed weldments. The effect of tritium on weldment toughness has not been measured until now. The data include tests on tritium-exposed weldments after aging for up to three years to measure the effect of increasing decay helium concentration on toughness. The results indicate that Type 304L stainless steel weldments have high fracture toughness and are resistant to tritium aging effects on toughness. For unexposed alloys, weldment fracture toughness was higher than base metal toughness. Tritium-exposed-and-aged base metals and weldments had lower toughness values than unexposed ones but still retained good toughness properties. In both base metals and weldments there was an initial reduction in fracture toughness after tritium exposure but little change in fracture toughness values with increasing helium content in the range tested. Fracture modes occurred by the dimpled rupture process in unexposed and tritium-exposed steels and welds. This corroborates further the resistance of Type 304L steel to tritium embrittlement. This report fulfills the requirements for the FY06 Level 3 milestone, TSR15.3 ''Issue summary report for tritium reservoir material aging studies'' for the Enhanced Surveillance Campaign (ESC). The milestone was in support of ESC L2-1866 Milestone-''Complete an annual Enhanced Surveillance stockpile aging assessment report to support the annual assessment process''.

  2. Fermilab | Tritium at Fermilab | Tritium released into the air and disposed

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

    of as solid waste Tritium released into the air and disposed of as solid waste Fermilab produces tritium as an expected byproduct of accelerator operations. The lab actively manages tritium, using and disposing of it in ways that pose no health or environmental threat. One of the ways that tritium is discharged from the Fermilab site is by releasing it into the air. This release occurs in various ways. Tritium in the form of water vapor is emitted into the air through ventilation systems

  3. Thermal Removal of Tritium from Concrete and Soil to Reduce Groundwater Impacts - 13197

    SciTech Connect (OSTI)

    Jackson, Dennis G.; Blount, Gerald C.; Wells, Leslie H.; Cardoso, Joao E.; Kmetz, Thomas F.; Reed, Misty L.

    2013-07-01

    Legacy heavy-water moderator operations at the Savannah River Site (SRS) have resulted in the contamination of equipment pads, building slabs, and surrounding soil with tritium. At the time of discovery the tritium had impacted the shallow (< 3-m) groundwater at the facility. While tritium was present in the groundwater, characterization efforts determined that a significant source remained in a concrete slab at the surface and within the associated vadose zone soils. To prevent continued long-term impacts to the shallow groundwater a CERCLA non-time critical removal action for these source materials was conducted to reduce the leaching of tritium from the vadose zone soils and concrete slabs. In order to minimize transportation and disposal costs, an on-site thermal treatment process was designed, tested, and implemented. The on-site treatment consisted of thermal detritiation of the concrete rubble and soil. During this process concrete rubble was heated to a temperature of 815 deg. C (1,500 deg. F) resulting in the dehydration and removal of water bound tritium. During heating, tritium contaminated soil was used to provide thermal insulation during which it's temperature exceeded 100 deg. C (212 deg. F), causing drying and removal of tritium. The thermal treatment process volatiles the water bound tritium and releases it to the atmosphere. The released tritium was considered insignificant based upon Clean Air Act Compliance Package (CAP88) analysis and did not exceed exposure thresholds. A treatability study evaluated the effectiveness of this thermal configuration and viability as a decontamination method for tritium in concrete and soil materials. Post treatment sampling confirmed the effectiveness at reducing tritium to acceptable waste site specific levels. With American Recovery and Reinvestment Act (ARRA) funding three additional treatment cells were assembled utilizing commercial heating equipment and common construction materials. This provided a total

  4. 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.

  5. Vanadium hydride deuterium-tritium generator

    DOE Patents [OSTI]

    Christensen, Leslie D.

    1982-01-01

    A pressure controlled vanadium hydride gas generator to provide deuterium-tritium gas in a series of pressure increments. A high pressure chamber filled with vanadium-deuterium-tritium hydride is surrounded by a heater which controls the hydride temperature. The heater is actuated by a power controller which responds to the difference signal between the actual pressure signal and a programmed pressure signal.

  6. 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.

  7. 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

  8. 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

  9. National Ignition Facility environmental protection systems

    SciTech Connect (OSTI)

    Mintz, J.M.; Reitz, T.C.; Tobin, M.T.

    1994-06-01

    The conceptual design of Environmental Protection Systems (EPS) for the National Ignition Facility (NIF) is described. These systems encompass tritium and activated debris handling, chamber, debris shield and general decontamination, neutron and gamma monitoring, and radioactive, hazardous and mixed waste handling. Key performance specifications met by EPS designs include limiting the tritium inventory to 300 Ci and total tritium release from NIF facilities to less than 10 Ci/yr. Total radiation doses attributable to NIF shall remain below 10 mrem/yr for any member of the general public and 500 mrem/yr for NIF staff. ALARA-based design features and operational procedures will, in most cases, result in much lower measured exposures. Waste minimization, improved cycle time and reduced exposures all result from the proposed CO2 robotic arm cleaning and decontamination system, while effective tritium control is achieved through a modern system design based on double containment and the proven detritiation technology.

  10. Progresses in tritium accident modelling in the frame of IAEA EMRAS II

    SciTech Connect (OSTI)

    Galeriu, D.; Melintescu, A.

    2015-03-15

    The assessment of the environmental impact of tritium release from nuclear facilities is a topic of interest in many countries. In the IAEA's Environmental Modelling for Radiation Safety (EMRAS I) programme, progresses for routine releases were done and in the EMRAS II programme a dedicated working group (WG 7 - Tritium Accidents) focused on the potential accidental releases (liquid and atmospheric pathways). The progresses achieved in WG 7 were included in a complex report - a technical document of IAEA covering both liquid and atmospheric accidental release consequences. A brief description of the progresses achieved in the frame of EMRAS II WG 7 is presented. Important results have been obtained concerning washout rate, the deposition on the soil of HTO and HT, the HTO uptake by leaves and the subsequent conversion to OBT (organically bound tritium) during daylight. Further needs of the processes understanding and the experimental efforts are emphasised.

  11. Tritium monitor with improved gamma-ray discrimination

    DOE Patents [OSTI]

    Cox, Samson A.; Bennett, Edgar F.; Yule, Thomas J.

    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.

  12. 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.

  13. DECOMMISSIONING THE HIGH PRESSURE TRITIUM LABORATORY AT LOS ALAMOS NATIONAL LABORATORY

    SciTech Connect (OSTI)

    Peifer, M.J.; Rendell, K.; Hearnsberger, D.W.

    2003-02-27

    In May 0f 2000, the Cerro Grande wild land fire burned approximately 48,000 acres in and around Los Alamos. In addition to the many buildings that were destroyed in the town site, many structures were also damaged and destroyed within the 43 square miles that comprise the Los Alamos National Laboratory (LANL). A special Act of Congress provided funding to remove Laboratory structures that were damaged by the fire, or that could be threatened by subsequent catastrophic wild land fires. The High Pressure Tritium Laboratory (HPTL) is located at Technical Area (TA) 33, building 86 in the far southeast corner of the Laboratory property. It is immediately adjacent to Bandelier National Park. Because it was threatened by both the Cerro Grande fire in 2000, and the 16,000- acre Dome fire in 1996, the former tritium processing facility was placed on the list of facilities scheduled for Decontamination and Decommissioning under the Cerro Grande Rehabilitation Project. The work was performed through the Facilities and Waste Operations (FWO) Division and is integrated with other Laboratory D&D efforts. The primary demolition contractor was Clauss Construction of San Diego, California. Earth Tech Global Environmental Services of San Antonio, Texas was sub-contracted to Clauss Construction, and provided radiological decontamination support to the project. Although the forty-seven year old facility had been in a state of safe-shutdown since operations ceased in 1990, a significant amount of tritium remained in the rooms where process systems were located. Tritium was the only radiological contaminant associated with this facility. Since no specific regulatory standards have been set for the release of volumetrically contaminated materials, concentration guidelines were derived in order to meet other established regulatory criteria. A tritium removal system was developed for this project with the goal of reducing the volume of tritium concentrated in the concrete of the building

  14. Microsoft Word - fact sheet Tritium 082814.docx

    National Nuclear Security Administration (NNSA)

    Tritium WHAT IS TRITIUM? Tritium is an isotope of hydrogen that occurs naturally in very small quantities. Hydrogen has three isotopes:  Protium Ordinary hydrogen with one proton and one electron in the atom. When two atoms of protium are combined with one atom of oxygen, water is created. Ordinary hydrogen comprises over 99.9 percent of all naturally occurring hydrogen.  Deuterium Sometimes called "heavy hydrogen," a non-radioactive isotope that has a neutron in the atom, in

  15. 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.

  16. Meeting Attendance - 34th Tritium Focus Group Meeting | Department of

    Office of Environmental Management (EM)

    Energy 4th Tritium Focus Group Meeting Meeting Attendance - 34th Tritium Focus Group Meeting Attendees to the 34th Tritium Focus Group Meeting held in Idaho Falls, ID on September 23-25, 2014. Meeting Attendance - 34th Tritium Focus Group Meeting (57.55 KB) More Documents & Publications Technological Assessment of Plasma Facing Components for DEMO Reactors Overview of tritium activity in Japan

  17. Secure Wireless Tritium Air Monitoring Cart Development | Department of

    Office of Environmental Management (EM)

    Energy Secure Wireless Tritium Air Monitoring Cart Development Secure Wireless Tritium Air Monitoring Cart Development Presentation from the 33rd Tritium Focus Group Meeting held in Aiken, South Carolina on April 22-24, 2014. Secure Wireless Tritium Air Monitoring Cart Development (4.14 MB) More Documents & Publications FY 2011 LDRD Report Tritium Detection Methods and Limitations FY 2008 LDRD Report

  18. 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.

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

    DOE Patents [OSTI]

    Jensen, George A.; Nelson, David A.; Molton, Peter M.

    1992-01-01

    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.

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

    Office of Environmental Management (EM)

    Behavior in Lead Lithium Eutectic (LLE) at Low Tritium Partial Pressure 33 rd Tritium Focus Group meeting, Savannah River National Laboratory, SC Masashi Shimada, Ph.D. Fusion Safety Program, Idaho National Laboratory, STIMS # INL/MS-14-31893| Savannah River National Laboratory, SC | April 25, 2014 Outlines 1. Motivation 2. Experimental apparatus 3. TMAP modeling 4. Experimental results 5. Modeling results 6. Future work M.Shimada | Tritium Focus Group meeting | SRNL, SC | April 25, 2014 2

  1. 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.

  2. 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...

  3. Fermilab | Tritium at Fermilab | Ferry Creek Results

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

    has been taken and analyzed. For samples in which a level of tritium above the limit of detection has been measured, the uncertainty of the measurement is indicated by an error...

  4. Tritium Issues in Next Step Devices

    SciTech Connect (OSTI)

    C.H. Skinner; G. Federici

    2001-09-05

    Tritium issues will play a central role in the performance and operation of next-step deuterium-tritium (DT) burning plasma tokamaks and the safety aspects associated with tritium will attract intense public scrutiny. The orders-of-magnitude increase in duty cycle and stored energy will be a much larger change than the increase in plasma performance necessary to achieve high fusion gain and ignition. Erosion of plasma-facing components will scale up with the pulse length from being barely measurable on existing machines to centimeter scale. Magnetic Fusion Energy (MFE) devices with carbon plasma-facing components will accumulate tritium by co-deposition with the eroded carbon and this will strongly constrain plasma operations. We report on a novel laser-based method to remove co-deposited tritium from carbon plasma-facing components in tokamaks. A major fraction of the tritium trapped in a co-deposited layer during the deuterium-tritium (DT) campaign on the Tokamak Fusion Test Reactor (TFTR) was released by heating with a scanning laser beam. This technique offers the potential for tritium removal in a next-step DT device without the use of oxidation and the associated deconditioning of the plasma-facing surfaces and expense of processing large quantities of tritium oxide. The operational lifetime of alternative materials such as tungsten has significant uncertainties due to melt layer loss during disruptions. Production of dust and flakes will need careful monitoring and minimization, and control and accountancy of the tritium inventory will be critical issues. Many of the tritium issues in Inertial Fusion Energy (IFE) are similar to MFE, but some, for example those associated with the target factory, are unique to IFE. The plasma-edge region in a tokamak has greater complexity than the core due to lack of poloidal symmetry and nonlinear feedback between the plasma and wall. Sparse diagnostic coverage and low dedicated experimental run time has hampered the

  5. Fermilab | Tritium at Fermilab | Indian Creek Results

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

    Indian Creek Results chart This chart (click chart for larger version) shows the levels of tritium in Indian Creek since November 2005, when our environmental monitoring program detected low levels of tritium in Indian Creek for the first time in its 35-year history, well below the federal water standards that Fermilab is required to meet. The detection limit is one picocurie per milliliter (see footnote below). Fermilab continues to monitor Indian Creek frequently and the results are displayed

  6. Vanadium hydride deuterium-tritium generator

    DOE Patents [OSTI]

    Christensen, L.D.

    1980-03-13

    A pressure controlled vanadium hydride gas generator was designed to provide deuterium-tritium gas in a series of pressure increments. A high pressure chamber filled with vanadium-deuterium-tritium hydride is surrounded by a heater which controls the hydride temperature. The heater is actuated by a power controller which responds to the difference signal between the actual pressure signal and a programmed pressure signal.

  7. 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.

  8. Recommended tritium surface contamination release guides

    SciTech Connect (OSTI)

    Johnson, J.R.; Draper, D.G.; Foulke, J.D.; Hafner, R.S.; Jalbert, R.A.; Kennedy, W.E.; Myers, D.S.; Strain, C.D. )

    1991-03-01

    This document was prepared to provide scientific basis for recommended changes in specific limits for tritium surface contamination in DOE Order 5480.11. A summary of the physical and biological characteristics of tritium has been provided that illustrate the unique nature of this radionuclide when compared to other pure beta emitters or to beta-gamma emitting radionuclides. This document is divided into nine sections. The introduction and the purpose and scope are addressed in Section 1.0 and Section 2.0, respectively. Section 3.0 contains recommended interpretation of terms used in this document. Section 4.0 addresses recommended methods for evaluating surface contamination. Biological and physical characteristics of tritium compounds are discussed in Section 5.0, as they relate to tritium radiotoxicity. Scenarios and dose calculations for selected, conservatively limiting cases of tritium intake are given and discussed in Section 6.0 and Section 7.0. Section 8.0 provides conclusions on the information given and recommendations for changes in the surface contamination limits for total tritium to 1 {times} 10{sup 6} dpm per 100 cm{sup 2}. 30 refs., 2 tabs.

  9. Guide to user facilities at the Lawrence Berkeley Laboratory

    SciTech Connect (OSTI)

    Not Available

    1984-04-01

    Lawrence Berkeley Laboratories' user facilities are described. Specific facilities include: the National Center for Electron Microscopy; the Bevalac; the SuperHILAC; the Neutral Beam Engineering Test Facility; the National Tritium Labeling Facility; the 88 inch Cyclotron; the Heavy Charged-Particle Treatment Facility; the 2.5 MeV Van de Graaff; the Sky Simulator; the Center for Computational Seismology; and the Low Background Counting Facility. (GHT)

  10. 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.