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 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.

  18. 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

  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. 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.

  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. 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

  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

    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

  2. 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.

  3. 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.

  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. 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.

  8. 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.

  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. 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

  14. Introduction Airborne Tritium Tritides

    Broader source: Energy.gov [DOE]

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

  15. Tritium 2013 Presentation

    Broader source: Energy.gov [DOE]

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

  16. 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.

  17. 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.

  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. 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.

  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 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 ...

  6. 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 ...

  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 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.

  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 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).

  19. 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.

  20. 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)

  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. 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

  5. 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

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  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, 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.

  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, 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.

  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. 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.

  3. 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.

  4. 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.

  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. 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)

  8. 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.

  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. Facility Floorplan

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

    facility floorplan Facility Floorplan

  15. 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.

  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. 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" ...

  19. 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.

  20. 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.

  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 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.

  3. 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

  4. 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.

  5. 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.

  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. 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.

  9. 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.

  10. 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.

  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. 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

  7. Recovery of tritium from tritiated molecules

    DOE Patents [OSTI]

    Swansiger, W.A.

    1984-10-17

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

  8. 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. 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.

  11. 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...

  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. 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 ...

  15. 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. ...

  16. 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...

  17. 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 ...

  18. 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...

  19. 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 ...

  20. 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 ...

  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. 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

  17. 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

  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. 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)

  8. 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.

  9. 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.

  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.

  11. Method and apparatus for controlling accidental releases of tritium

    DOE Patents [OSTI]

    Galloway, T.R.

    1980-04-01

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

  12. Radioactive waste tank ventilation system incorporating tritium control

    SciTech Connect (OSTI)

    Rice, P.D.

    1997-08-01

    This paper describes the development of a ventilation system for radioactive waste tanks at the U.S. Department of Energy`s (DOE) Hanford Site in Richland, Washington. The unique design of the system is aimed at cost-effective control of tritiated water vapor. The system includes recirculation ventilation and cooling for each tank in the facility and a central exhaust air clean-up train that includes a low-temperature vapor condenser and high-efficiency mist eliminator (HEME). A one-seventh scale pilot plant was built and tested to verify predicted performance of the low-temperature tritium removal system. Tests were conducted to determine the effectiveness of the removal of condensable vapor and soluble and insoluble aerosols and to estimate the operating life of the mist eliminator. Definitive design of the ventilation system relied heavily on the test data. The unique design features of the ventilation system will result in far less release of tritium to the atmosphere than from conventional high-volume dilution systems and will greatly reduce operating costs. NESHAPs and TAPs NOC applications have been approved, and field construction is nearly complete. Start-up is scheduled for late 1996. 3 refs., 4 figs., 2 tabs.

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

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

    The Office of Enforcement and Oversight (Independent Oversight), within the Office of Health, Safety and Security (HSS), conducted an independentreview of preparedness for severe ...

  14. Tritium-helium effects in metals

    SciTech Connect (OSTI)

    Caskey, G.R.

    1985-09-01

    Investigations of helium effects in metals at the Savannah River Laboratory have been carried out by introducing helium by radioactive decay of tritium. This process does not create concurrent radiation damage, such as accompanies ion implantation and (n,..cap alpha..) reactions. The process has its own peculiarities, however, which partially mask and interact with the helium effect of interest. The distribution and local concentration of helium and tritium, which are responsible for changes in mechanical properties and fracture mode, are controlled by the large difference in solubility and diffusivity between the two atoms and by their differing interaction energies with lattice defects, impurities, and internal boundaries. Furthermore, in all investigations with helium generated from tritium decay, some tritium and deuterium are always present. Consequently, property changes include tritium-helium interaction effects to some extent. Results of investigations with several austenitic stainless steels, Armco iron, and niobium single crystals illustrate the variety of phenomena and some of the complex interactions that can be encountered.

  15. Tritium-helium effects in metals

    SciTech Connect (OSTI)

    Caskey, G.R. Jr.

    1985-01-01

    Investigations of helium effects in metals at the Savannah River Laboratory have been carried out by introducing helium by radioactive decay of tritium. This process does not create concurrent radiation damage, such as accompanies ion implantation and (n,..cap alpha..) reactions. The process has its own peculiarities, however, which partially mask and interact with the helium effect of interest. The distribution and local concentration of helium and tritium, which are responsible for changes in mechanical properties and fracture mode, are controlled by the large difference in solubility and diffusivity between the two atoms and by their differing interaction energies with lattice defects, impurities, and internal boundaries. Furthermore, in all investigations with helium generated from tritium decay, some tritium and deuterium are always present. Consequently, property changes include tritium-helium interaction effects to some extent. Results of investigations with several austenitic stainless steels, Armco iron, and niobium single crystals illustrate the variety of phenomena and some of the complex interactions that can be encountered.

  16. Tritium R&D at AECL Selected Topics

    Office of Environmental Management (EM)

    ... within 5C * Sample: * 10 mm dia. * 50 mm lg. * HT flow inner * Swept annulus * Measure ... UNRESTRICTED ILLIMIT Direct Tritium battery Immobilized Tritium Layers 16 Ti TiT 2 ...

  17. Tritium diffusion in lithium oxide solid breeder materials

    SciTech Connect (OSTI)

    Shearer, J.A.; Tam, S.W.; Johnson, C.E.

    1983-01-01

    A review of data of tritium diffusion in Li/sub 2/O is presented. Also diffusion coefficients in Li/sub 2/O of tritium, lithium, oxygen, hydrogen, and deuterium are given. (MOW)

  18. Advancement Of Tritium Powered Betavoltaic Battery Systems

    SciTech Connect (OSTI)

    Staack, G.; Gaillard, J.; Hitchcock, D.; Peters, B.; Colon-Mercado, H.; Teprovich, J.; Coughlin, J.; Neikirk, K.; Fisher, C.

    2015-10-14

    Due to their decades-long service life and reliable power output under extreme conditions, betavoltaic batteries offer distinct advantages over traditional chemical batteries, especially in applications where frequent battery replacement is hazardous, or cost prohibitive. Although many beta emitting isotopes exist, tritium is considered ideal in betavoltaic applications for several reasons: 1) it is a “pure” beta emitter, 2) the beta is not energetic enough to damage the semiconductor, 3) it has a moderately long half-life, and 4) it is readily available. Unfortunately, the widespread application of tritium powered betavoltaics is limited, in part, by their low power output. This research targets improving the power output of betavoltaics by increasing the flux of beta particles to the energy conversion device (the p-n junction) through the use of low Z nanostructured tritium trapping materials.

  19. Overview of AECL’s Tritium Compatible Electrolyser Program

    Broader source: Energy.gov [DOE]

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

  20. Tritium R&D at AECL Selected Topics

    Office of Energy Efficiency and Renewable Energy (EERE)

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

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

    Broader source: Energy.gov [DOE]

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

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

    Office of Energy Efficiency and Renewable Energy (EERE)

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

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

    Broader source: Energy.gov [DOE]

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

  4. Meeting Attendance- 33rd Tritium Focus Group Meeting

    Broader source: Energy.gov [DOE]

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

  5. Meeting Attendance- 36th Tritium Focus Group Meeting

    Broader source: Energy.gov [DOE]

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

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

    Broader source: Energy.gov [DOE]

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

  7. Tritium Gas Stream Scrubbing using In-situ Reactive Materials

    Office of Energy Efficiency and Renewable Energy (EERE)

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

  8. Tritium Science and Technology at Atomic Weapons Establishment (AWE)

    Broader source: Energy.gov [DOE]

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

  9. Overview of Tritium Betavoltaic Power for Micro Sensors

    Broader source: Energy.gov [DOE]

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

  10. Tritium Aging Effects in Palladium on Kieselguhr

    SciTech Connect (OSTI)

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

    1998-10-01

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

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

    Broader source: Energy.gov [DOE]

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

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

    SciTech Connect (OSTI)

    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 River 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

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

    Broader source: Energy.gov [DOE]

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

  14. Microsoft PowerPoint - Tritium Design Practice.ppt

    Office of Environmental Management (EM)

    LET'S COMPARE TRITIUM DESIGN PRACTICES ACROSS THE DOE COMPLEX X Steve Xiao 2 GENERAL REFERENCES DOE-HDBK-1129-2008, "DOE Handbook Tritium Handling and Safe Storage", 2008 William W Weaver et al, DOE/EH-0417, Technical Notice Issue No 94-01, "Guidelines for Valves in Tritium Service", 1994 F Mannone, Springer, Editor, "Safety in Tritium Handling Technology", Kluwer Academic Publishers, 1993 International Atomic Energy Agency, "Safe Handling of Tritium: Review of

  15. Small system for tritium accelerator mass spectrometry

    DOE Patents [OSTI]

    Roberts, Mark L.; Davis, Jay C.

    1993-01-01

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

  16. Small system for tritium accelerator mass spectrometry

    DOE Patents [OSTI]

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

    1993-02-23

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

  17. DEVELOPMENT OF THE BULK TRITIUM SHIPPING PACKAGING

    SciTech Connect (OSTI)

    Blanton, P.; Eberl, K.

    2008-09-14

    A new radioactive shipping packaging for transporting bulk quantities of tritium, the Bulk Tritium Shipping Package (BTSP), has been designed for the Department of Energy (DOE) as a replacement for a package designed in the early 1970s. This paper summarizes significant design features and describes how the design satisfies the regulatory safety requirements of the Code of Federal Regulations and the International Atomic Energy Agency. The BTSP design incorporates many improvements over its predecessor by implementing improved testing, handling, and maintenance capabilities, while improving manufacturability and incorporating new engineered materials. This paper also discusses the results from testing of the BTSP to 10 CFR 71 Normal Conditions of Transport and Hypothetical Accident Condition events. The programmatic need of the Department of Energy (DOE) to ship bulk quantities of tritium has been satisfied since the late 1970s by the UC-609 shipping package. The current Certificate of Conformance for the UC-609, USA/9932/B(U) (DOE), will expire in late 2011. Since the UC-609 was not designed to meet current regulatory requirements, it will not be recertified and thereby necessitates a replacement Type B shipping package for continued DOE tritium shipments in the future. A replacement tritium packaging called the Bulk Tritium Shipping Package (BTSP) is currently being designed and tested by Savannah River National Laboratory (SRNL). The BTSP consists of two primary assemblies, an outer Drum Assembly and an inner Containment Vessel Assembly (CV), both designed to mitigate damage and to protect the tritium contents from leaking during the regulatory Hypothetical Accident Condition (HAC) events and during Normal Conditions of Transport (NCT). During transport, the CV rests on a silicone pad within the Drum Liner and is covered with a thermal insulating disk within the insulated Drum Assembly. The BTSP packaging weighs approximately 500 lbs without contents and is 50

  18. In-bed measurement of tritium loading in process metal hydride beds

    SciTech Connect (OSTI)

    Nobile, A.

    1988-01-01

    The Replacement Tritium Facility at Savannah River Plant will make extensive use of metal hydride technology for the storage, pumping, isotopic separation, and compression of hydrogen isotopes. Two options were considered for routine accountability of tritium stored in metal hydride beds. One option was to use standard P-V-T-mass spectrometry techniques after desorption of storage beds to tanks of known volume. The second option was to develop a technique for direct measurement of bed loading. It was thought that such a technique would be more rapid and would account for heel, although some accuracy would be lost.The static nitrogen and flowing nitrogen methods were considered for this option. The flowing nitrogen method was eventually selected because it was insensitive to bed physical properties and isotopic gas composition, as well as being more accurate and easier to automate.

  19. Apparatus to recover tritium from tritiated molecules

    DOE Patents [OSTI]

    Swansiger, William A. (Livermore, CA)

    1988-01-01

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

  20. Tritium High Vacuum Pump Test Plan

    Office of Environmental Management (EM)

    DOE-HDBK-1129-2008 December 2008 DOE HANDBOOK TRITIUM HANDLING AND SAFE STORAGE U.S. Department of Energy AREA SAFT Washington, D.C. 20585 DISTRIBUTION STATEMENT A. Approved for public release; distribution is unlimited. TS DOE-HDBK-1129-2008 ii This page is intentionally blank. DOE-HDBK-1129-2008 iii TABLE OF CONTENTS SECTION PAGE FOREWORD................................................................................................................................ ix ACRONYMS

  1. Recovery of tritium from tritiated molecules

    DOE Patents [OSTI]

    Swansiger, William A. (Livermore, CA)

    1987-01-01

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

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

    Office of Environmental Management (EM)

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

  3. DOE - Office of Legacy Management -- Rulsion Tritium Transport...

    Office of Legacy Management (LM)

    September 2007 pdficon Tritium Transport Model Comments and Responses Colorado Oil and Gas Conservation Commission Colorado Department of Public Health and Environment ...

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

    Office of Environmental Management (EM)

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

  5. Characterization of Tritium Breeding Ratio and Energy Multiplication...

    Office of Scientific and Technical Information (OSTI)

    Factor of Lithium-based Ternary Alloys in IFE Blankets Citation Details In-Document Search Title: Characterization of Tritium Breeding Ratio and Energy Multiplication Factor of ...

  6. Overview of Japanese activities on tritium research for fusion...

    Office of Environmental Management (EM)

    (NIRS) TritiumMaterial Interaction - Plasma Facing Materials - Structural Materials ... Society of Japan The Japan Society of Plasma Science and Nuclear Fusion Research ...

  7. TRITIUM BARRIER MATERIALS AND SEPARATION SYSTEMS FOR THE NGNP

    SciTech Connect (OSTI)

    Sherman, S; Thad Adams, T

    2008-07-17

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

  8. Behavior of tritium permeation induced by water corrosion of...

    Office of Scientific and Technical Information (OSTI)

    induced by water corrosion of alpha iron around room temperature Citation Details In-Document Search Title: Behavior of tritium permeation induced by water corrosion of ...

  9. Microsoft PowerPoint - Tritium Design Practice.ppt

    Office of Environmental Management (EM)

    Publishers, 1993 International Atomic Energy Agency, ... Floor plan available space Experiences lessens learnt 10 ... EXAMPLE 1 - TRITIUM INSTRUMENT TEST STATION 15 USE OF GAS ...

  10. Electron Proton Hydrogen Deuterium Tritium Neutron Fusion Basics

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

    Hydrogen Deuterium Tritium Neutron Fusion Basics Throughout history, the way in which the sun and stars produce their energy remained a mystery. During the 20th century, scientists ...

  11. Low tritium partial pressure permeation system for mass transport...

    Office of Scientific and Technical Information (OSTI)

    Low tritium partial pressure permeation system for mass transport measurement in lead lithium eutectic Citation Details In-Document Search This content will become publicly ...

  12. Tritium release from SS316 under vacuum condition

    SciTech Connect (OSTI)

    Torikai, Y.; Penzhorn, R.D.

    2015-03-15

    The plasma facing surface of the ITER vacuum vessel, partly made of low carbon austenitic stainless steel type 316L, will incorporate tritium during machine operation. In this paper the kinetics of tritium release from stainless steel type 316 into vacuum and into a noble gas stream are compared and modelled. Type 316 stainless steel specimens loaded with tritium either by exposure to 1.2 kPa HT at 573 K or submersion into liquid HTO at 298 K showed characteristic thin surface layers trapping tritium in concentrations far higher than those determined in the bulk. The evolution of the tritium depth profile in the bulk during heating under vacuum was non-discernible from that of tritium liberated into a stream of argon. Only the relative amount of the two released tritium-species, i.e. HT or HTO, was different. Temperature-dependent depth profiles could be predicted with a one-dimensional diffusion model. Diffusion coefficients derived from fitting of the tritium release into an evacuated vessel or a stream of argon were found to be (1.4 ± 1.0)*10{sup -7} and (1.3 ± 0.9)*10{sup -9} cm{sup 2}/s at 573 and 423 K, respectively. Polished surfaces on type SS316 stainless steel inhibit considerably the thermal release rate of tritium.

  13. 2012 ACCOMPLISHMENTS - TRITIUM AGING STUDIES ON STAINLESS STEELS

    SciTech Connect (OSTI)

    Morgan, M.

    2013-01-31

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

  14. Tritium Formation and Mitigation in High-Temperature Reactors

    SciTech Connect (OSTI)

    Piyush Sabharwall; Carl Stoots

    2012-10-01

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

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

    SciTech Connect (OSTI)

    Piyush Sabharwall; Carl Stoots; Hans A. Schmutz

    2013-03-01

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

  16. EFFECTS OF TRITIUM GAS EXPOSURE ON EPDM ELASTOMER

    SciTech Connect (OSTI)

    Clark, E.

    2009-12-11

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

  17. Continuous monitors for tritium in sodium coolant and cover gas...

    Office of Scientific and Technical Information (OSTI)

    Continuous monitors for tritium in the sodium coolant and the cover gas of a fast breeder reactor have been ... Language: English Subject: N77500* --Reactors--Power Reactors, ...

  18. Preliminary safety assessment for an IFE target fabrication facility

    SciTech Connect (OSTI)

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

    2000-10-13

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

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

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

    EIS-0288-S1: Production of Tritium in a Commercial Light Water Reactor (CLWR) Tritium ... February 24, 2016 EIS-0288-S1: Final Supplemental Environmental Impact Statement August ...

  20. New Tritium Source On Line at DOE's Savannah River Site | National...

    National Nuclear Security Administration (NNSA)

    Tritium Source On Line at DOE's Savannah River Site February 27, 2007 New Tritium Source On Line at DOE's Savannah River Site

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

    Office of Environmental Management (EM)

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

  2. Production of {sup 4}He, {sup 3}He, and tritium from Be irradiated in FFTF-MOTA-2B

    SciTech Connect (OSTI)

    Greenwood, L.R.

    1998-03-01

    The production of {sup 4}He, {sup 3}He, and tritium has been calculated for beryllium irradiated in the Materials Open Test Assembly (MOTA)-2B experiment in the Fast Flux Test Facility (FFTF). Reaction rates were based on adjusted neutron spectra determined from reactor dosimetry measurements at seven different elevations in the irradiation assembly. Equations are given so that gas production, dpa, and neutron fluences can be calculated for any specific elevation in the MOTA-2B assembly.

  3. Radiological Control Programs for Special Tritium Compounds

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

    84-2004 SEPTEMBER 2004 CHANGE NOTICE NO. 1 Date June 2006 DOE HANDBOOK RADIOLOGICAL CONTROL PROGRAMS FOR SPECIAL TRITIUM COMPOUNDS U.S. Department of Energy AREA OCSH Washington, D.C. 20585 DISTRIBUTION STATEMENT A. Approved for public release; distribution is unlimited. NOT MEASUREMENT SENSITIVE ii Table of Changes Page Change 67 (near bottom) In row 1, column 2 of the table titled "dosimetric properties" 6 mrem was changed to 6 x 10 -2 mrem Available on the Department of Energy

  4. Radiological Control Programs for Special Tritium Compounds

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

    DOE.F 1325.8 (08-93) United States Government Department of Energy memorandum DATE: May 11, 2006 REPLY TO EH-52:JRabovsky:3-2 135 ATTN OF: APPROVAL OF CHANGE NOTICE 1 TO DEPARTMENT OF ENERGY (DOE) SUBJECT. HANDBOOK 1184-2004, RADIOLOGICAL CONTROL PROGRAMS FOR SPECIAL TRITIUM COMPOUNDS TO: Dennis Kubicki, EH-24 Technical Standards Manager This memorandum forwards the subject Change Notice 1 to DOE Handbook, DOE- HDBK- 1184-2004, which has approved for publication and distribution. The change to

  5. Microsoft Word - Cooper_Tritium.final.doc

    Office of Legacy Management (LM)

    PROCEEDINGS, TOUGH Symposium 2006 Lawrence Berkeley National Laboratory, Berkeley, California, May 15-17, 2006 - 1 - TRITIUM TRANSPORT THROUGH A LOW-PERMEABILITY NATURAL GAS RESERVOIR Clay A. Cooper 1 , Ming Ye 2 , and Jenny Chapman 2 Desert Research Institute 1 2215 Raggio Pkwy Reno, Nevada, 89512, USA E-mail: clay@dri.edu 2 755 E. Flamingo Road Las Vegas, NV 89119, USA ABSTRACT The U.S. Department of Energy and its predecessor agencies conducted a program in the 1960s and 1970s to evaluate

  6. Tritium labeling of organic compounds deposited on porous structures

    DOE Patents [OSTI]

    Ehrenkaufer, Richard L. E.; Wolf, Alfred P.; Hembree, Wylie C.

    1979-01-01

    An improved process for labeling organic compounds with tritium is carried out by depositing the selected compound on the extensive surface of a porous structure such as a membrane filter and exposing the membrane containing the compound to tritium gas activated by the microwave discharge technique. The labeled compound is then recovered from the porous structure.

  7. Confinement and heating of a deuterium-tritium plasma

    SciTech Connect (OSTI)

    Hawryluk, R. J.; Adler, H.; Alling, P.; Synakowski, E.

    1994-03-01

    The Tokamak Fusion Test Reactor (TFTR) has performed initial high-power experiments with the plasma fueled by deuterium and tritium to nominally equal densities. Compared to pure deuterium plasmas, the energy stored in the electron and ions increased by ~20%. These increases indicate improvements in confinement associated with the use of tritium and possibly heating of electrons by α-particles.

  8. Microsoft PowerPoint - Electrolytic T Extraction in Molten Li-LiT_2.pptx

    Office of Environmental Management (EM)

    Electrolytic Tritium Extraction in Molten Li-LiT Luke Olson Brenda L. García-Díaz Hector Colon-Mercado Joe Teprovich Dave Babineau Savannah River National Laboratory Fall 2015 Tritium Focus Group Meeting November 3-5, 2015 SRNL-STI-2015-00605 This presentation does not contain any proprietary, confidential, or otherwise restricted information LiT Electrolysis Options LiT Electrolysis Maroni Process (Baseline Option) Improve Liquid-Liquid Extraction & Electrolysis Process Intensification

  9. Organically bound tritium analysis in environmental samples

    SciTech Connect (OSTI)

    Baglan, N.; Cossonnet, C.; Fournier, M.; Momoshima, N.; Ansoborlo, E.

    2015-03-15

    Organically bound tritium (OBT) has become of increased interest within the last decade, with a focus on its behaviour and also its analysis, which are important to assess tritium distribution in the environment. In contrast, there are no certified reference materials and no standard analytical method through the international organization related to OBT. In order to resolve this issue, an OBT international working group was created in May 2012. Over 20 labs from around the world participated and submitted their results for the first intercomparison exercise results on potato (Sep 2013). The samples, specially-prepared potatoes, were provided in March 2013 to each participant. Technical information and results from this first exercise are discussed here for all the labs which have realised the five replicates necessary to allow a reliable statistical treatment. The results are encouraging as the increased number of participating labs did not degrade the observed dispersion of the results for a similar activity level. Therefore, the results do not seem to depend on the analytical procedure used. From this work an optimised procedure can start to be developed to deal with OBT analysis and will guide subsequent planned OBT trials by the international group.

  10. Derivation of dose conversion factors for tritium

    SciTech Connect (OSTI)

    Killough, G. G.

    1982-03-01

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

  11. Nuclear Facilities Production Facilities

    National Nuclear Security Administration (NNSA)

    Facilities Production Facilities Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. Department of Energy's National Nuclear Security Administration under contract DE-AC04-94AL85000. Sand 2011-4582P. ENERGY U.S. DEPARTMENT OF Gamma Irradiation Facility (GIF) The GIF provides test cells for the irradiation of experiments with high-intensity gamma ray sources. The main features

  12. Facility Environmental Vulnerability Assessment

    SciTech Connect (OSTI)

    Van Hoesen, S.D.

    2001-07-09

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

  13. Tritium and neutron measurements from deuterated Pd-Si

    SciTech Connect (OSTI)

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

    1991-05-10

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

  14. Tritium and neutron measurements from deuterated Pd-Si

    SciTech Connect (OSTI)

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

    1990-01-01

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

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

    SciTech Connect (OSTI)

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

    2010-10-14

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

  16. Facility effluent monitoring plan for the 324 Facility

    SciTech Connect (OSTI)

    1994-11-01

    The 324 Facility [Waste Technology Engineering Laboratory] in the 300 Area primarily supports the research and development of radioactive and nonradioactive waste vitrification technologies, biological waste remediation technologies, spent nuclear fuel studies, waste mixing and transport studies, and tritium development programs. All of the above-mentioned programs deal with, and have the potential to, release hazardous and/or radioactive material. The potential for discharge would primarily result from (1) conducting research activities using the hazardous materials, (2) storing radionuclides and hazardous chemicals, and (3) waste accumulation and storage. This report summarizes the airborne and liquid effluents, and the results of the Facility Effluent Monitoring Plan (FEMP) determination for the facility. The complete monitoring plan includes characterizing effluent streams, monitoring/sampling design criteria, a description of the monitoring systems and sample analysis, and quality assurance requirements.

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

    DOE Patents [OSTI]

    Anderson, David F.

    1984-01-01

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

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

    DOE Patents [OSTI]

    Anderson, D.F.

    1981-01-27

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

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

    SciTech Connect (OSTI)

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

    1999-06-01

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

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

    SciTech Connect (OSTI)

    1994-12-05

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

  1. Cryogenic tritium-hydrogen-deuterium and deuterium-tritium layer implosions with high density carbon ablators in near-vacuum hohlraums

    SciTech Connect (OSTI)

    Meezan, N. B. Hopkins, L. F. Berzak; Pape, S. Le; Divol, L.; MacKinnon, A. J.; Döppner, T.; Ho, D. D.; Jones, O. S.; Khan, S. F.; Ma, T.; Milovich, J. L.; Pak, A. E.; Ross, J. S.; Thomas, C. A.; Benedetti, L. R.; Bradley, D. K.; Celliers, P. M.; Clark, D. S.; Field, J. E.; Haan, S. W.; and others

    2015-06-15

    High Density Carbon (or diamond) is a promising ablator material for use in near-vacuum hohlraums, as its high density allows for ignition designs with laser pulse durations of <10 ns. A series of Inertial Confinement Fusion (ICF) experiments in 2013 on the National Ignition Facility [Moses et al., Phys. Plasmas 16, 041006 (2009)] culminated in a deuterium-tritium (DT) layered implosion driven by a 6.8 ns, 2-shock laser pulse. This paper describes these experiments and comparisons with ICF design code simulations. Backlit radiography of a tritium-hydrogen-deuterium (THD) layered capsule demonstrated an ablator implosion velocity of 385 km/s with a slightly oblate hot spot shape. Other diagnostics suggested an asymmetric compressed fuel layer. A streak camera-based hot spot self-emission diagnostic (SPIDER) showed a double-peaked history of the capsule self-emission. Simulations suggest that this is a signature of low quality hot spot formation. Changes to the laser pulse and pointing for a subsequent DT implosion resulted in a higher temperature, prolate hot spot and a thermonuclear yield of 1.8 × 10{sup 15} neutrons, 40% of the 1D simulated yield.

  2. Tritium Formation and Mitigation in High Temperature Reactors

    SciTech Connect (OSTI)

    Piyush Sabharwall; Carl Stoots

    2012-08-01

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

  3. On-line tritium production monitor

    DOE Patents [OSTI]

    Mihalczo, J.T.

    1993-11-23

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

  4. Raman spectroscopy at the tritium laboratory Karlsruhe

    SciTech Connect (OSTI)

    Schloesser, M.; Bornschein, B.; Fischer, S.; Kassel, F.; Rupp, S.; Sturm, M.; James, T.M.; Telle, H.H.

    2015-03-15

    Raman spectroscopy is employed successfully for analysis of hydrogen isotopologues at the Tritium Laboratory Karlsruhe (TLK). Raman spectroscopy is based on the inelastic scattering of photons off molecules. Energy is transferred to the molecules as rotational/vibrational excitation being characteristic for each type of molecule. Thus, qualitative analysis is possible from the Raman shifted light, while quantitative information can be obtained from the signal intensities. After years of research and development, the technique is now well-advanced providing fast (< 10 s), precise (< 0.1%) and true (< 3%) compositional analysis of gas mixtures of hydrogen isotopologues. In this paper, we summarize the recent achievements in the further development on this technique, and the various applications for which it is used at TLK. Raman spectroscopy has evolved as a versatile, highly accurate key method for quantitative analysis complementing the port-folio of analytic techniques at the TLK.

  5. On-line tritium production monitor

    DOE Patents [OSTI]

    Mihalczo, John T.

    1993-01-01

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

  6. Ignition of deuterium-tritium fuel targets

    DOE Patents [OSTI]

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

    1991-08-27

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

  7. Tritium Migration Analysis Program Version 4

    Energy Science and Technology Software Center (OSTI)

    1991-06-12

    TMAP4 was developed as a safety analysis code, mainly to analyze tritium retention and loss in fusion reactor structures and systems during normal operational and accident conditions. It incorporates one-dimensional thermal and mass-diffusive transport and trapping calculations through structures and zero dimensional fluid transport between enclosures and across the interface between enclosures and structures. Diffusion structures may be linked together with other structures, and multiple structures may interact with an enclosure. A key feature ismore » the ability to input problem definition parameters as constants, interpolation tables, or FORTRAN equations. The code is specifically intended for use under a DOS operating system on PC type minicomputers, but it has also been run successfully on workstations and mainframe computer systems. Use of the equation-input feature requires access to a FORTRAN-77 compiler, and a linker program is required.« less

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

    SciTech Connect (OSTI)

    Senor, David J.

    2013-10-30

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

  9. 34th Tritium Focus Group Meeting, Idaho National Laboratory, Idaho Falls,

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

    September 23-25, 2014 | Department of Energy 4th Tritium Focus Group Meeting, Idaho National Laboratory, Idaho Falls, September 23-25, 2014 34th Tritium Focus Group Meeting, Idaho National Laboratory, Idaho Falls, September 23-25, 2014 34th Tritium Focus Group Meeting, Idaho National Laboratory, Idaho Falls, September 23-25, 2014 The Tritium Focus Group (TFG), is a long standing DOE Working Group, whose purpose is to promote cost-effective improvements in tritium safety, handling,

  10. Tritium removal from tritiated water by organic functionalized SBA-15

    SciTech Connect (OSTI)

    Taguchi, A.; Kato, Y.; Akai, R.; Torikai, Y.; Matsuyama, M.

    2015-03-15

    The recovery of tritium from tritiated water is important for reducing tritium emissions to the environment and for recycling tritium. Meso-porous silicas (SBA-15) were modified by -COOH, -SO{sub 3}H and -NH{sub 2} groups and their tritium adsorption ability from tritiated water under solid-liquid sorption was investigated. The adsorption abilities and separation factor of organic functionalized SBAs were comparable to those of bare SBA. The desorption of water from bare SBA and -COOH functionalized SBA were studied by Fourier transform infra-red spectroscopy using D{sub 2}O as a probe molecule. An interaction was observed for D{sub 2}O with -COOH group where the hydrogen bonds became weaker than D{sub 2}O with bare SBA. (authors)

  11. NNSA Marks Major Milestone For Tritium Production | National...

    National Nuclear Security Administration (NNSA)

    Tritium is a radioactive form of hydrogen gas that is an integral component in the nuclear weapons stockpile. It must be replenished in weapons periodically because it has a ...

  12. Fixation of tritium in a highly stable polymer form

    DOE Patents [OSTI]

    Steinberg, Meyer; Colombo, Peter; Pruzansky, Jacob

    1977-01-01

    A method for the fixation of tritium comprising reacting tritiated water with calcium carbide to produce calcium hydroxide and tritiated acetylene, polymerizing the acetylene, and then incorporating the polymer in a solidifying matrix.

  13. Record of Decision, Tritium Supply and Recycling Programmatic...

    Office of Environmental Management (EM)

    ... the tritium in pure form for its use in nuclear weapons. ... would include an interim spent fuel storage building, a waste ... for the assessment panels who did not stand to gain ...

  14. Tritium gettering from air with hydrogen uranyl phosphate

    SciTech Connect (OSTI)

    Souers, P.C.; Uribe, F.S.; Stevens, C.G.; Tsugawa, R.T.

    1985-01-01

    Hydrogen uranyl phosphate (HUP), a solid proton electrolyte, getters tritium gas and water vapor from air by DC electrical action. We have reduced the formation of residual tritiated water to less than 2%, and demonstrated that HUP can clean a 5.5 m/sup 3/ working glove box. Data are presented to illustrate the parameters of the gettering and a model is derived. Two other tritium gettering electrolytes have been discovered. 9 refs., 5 figs., 3 tabs.

  15. ARM - SGP Extended Facility

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

    Extended Facility SGP Related Links Virtual Tour Facilities and Instruments Central Facility Boundary Facility Extended Facility Intermediate Facility Radiometric Calibration...

  16. ARM - SGP Intermediate Facility

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

    Intermediate Facility SGP Related Links Virtual Tour Facilities and Instruments Central Facility Boundary Facility Extended Facility Intermediate Facility Radiometric Calibration...

  17. ARM - SGP Central Facility

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

    Central Facility SGP Related Links Virtual Tour Facilities and Instruments Central Facility Boundary Facility Extended Facility Intermediate Facility Radiometric Calibration...

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

    SciTech Connect (OSTI)

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

    2000-06-01

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

  19. Pre-operational safety appraisal Tritiated Scrap Recovery Facility, Mound facility

    SciTech Connect (OSTI)

    Dauby, J.J.; Flanagan, T.M.; Metcalf, L.W.; Rhinehammer, T.B.

    1996-07-01

    The purpose of this report is to identify, assess, and document the hazards which are associated with the proposed operation of the Tritiated Scrap Recovery Facility at Mound Facility. A Pre-operational Safety Appraisal is a requirement as stated in Department of Energy Order 5481.1, Safety Analysis and Review System. The operations to be conducted in the new Tritiated Scrap Waste Recovery Facility are not new, but a continuation of a prime mission of Mound`s i.e. recovery of tritium from waste produced throughout the DOE complex. The new facility is a replacement of an existing process started in the early 1960`s and incorporates numerous design changes to enhance personnel and environmental safety. This report also documents the safety of a one time operation involving the recovery of tritium from material obtained by the Department of Energy from the State of Arizona. This project will involve the processing of 240,000 curies of tritium contained in glass ampoules that were to be used in items such as luminous dial watches. These were manufactured by the now defunct American Atomics Corporation, Tucson, Arizona.

  20. Huntington Resource Recovery Facility Biomass Facility | Open...

    Open Energy Info (EERE)

    Resource Recovery Facility Biomass Facility Jump to: navigation, search Name Huntington Resource Recovery Facility Biomass Facility Facility Huntington Resource Recovery Facility...

  1. Wheelabrator Sherman Energy Facility Biomass Facility | Open...

    Open Energy Info (EERE)

    Sherman Energy Facility Biomass Facility Jump to: navigation, search Name Wheelabrator Sherman Energy Facility Biomass Facility Facility Wheelabrator Sherman Energy Facility Sector...

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

    SciTech Connect (OSTI)

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

    1994-06-01

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

  3. Key technologies for tritium storage bed development

    SciTech Connect (OSTI)

    Yu, S.H.; Chang, M.H.; Kang, H.G.; Chung, D.Y.; Oh, Y.H.; Jung, K.J.; Chung, H.; Koo, D.; Sohn, S.H.; Song, K.M.

    2015-03-15

    ITER Storage and Delivery System (SDS) is a complex system involving tens of storage beds. The most important SDS getter bed will be used for the absorption and desorption of hydrogen isotopes in accordance with the fusion fuel cycle scenario. In this paper the current status concerning research/development activities for the optimal approach to the final SDS design is introduced. A thermal analysis is performed and discussed on the aspect of heat losses considering whether the reflector and/or the feed-through is present or not. A thermal hydraulic simulation shows that the presence of 3 or 4 reflectors minimize the heat loss. Another important point is to introduce the real-time gas analysis in the He{sup 3} collection system. In this study 2 independent strength methods based on gas chromatography and quadruple mass spectrometer for one and on a modified self-assaying quadruple mass spectrometer for the second are applied to separate the hydrogen isotopes in helium gas. Another issue is the possibility of using depleted uranium getter material for the storage of hydrogen isotopes, especially of tritium.

  4. Three tritium systems test assembly (TSTA) off-loop experiments

    SciTech Connect (OSTI)

    Talcott, C.L.; Anderson, J.L.; Carlson, R.V.; Coffin, D.O.; Walthers, C.R.; Hamerdinger, D.; Binning, K.; Trujillo, R.D.; Moya, J.S.; Hayashi, T.; Okuno, K.; Yamanishi, T.

    1993-11-01

    This report contains the results from three different experiments. Experiment one was initiated to establish the possibility of using a soft elastomer in ITER (International Thermonuclear Experimental Reactor) applications. Used in this application, the sealing material is anticipated to be in tritium at pressures in the range of 1 {times} 10{sup {minus}3} torr for many years. Here two O-ring valve seals each of Viton-A, Buna-N, and EDPM were exposed to 1, 40, or 400 torr of tritium while being cycled open and closed approximately 11,500 times in 192 days. EDPM is the least susceptible to damage from the tritium. Both Buna-N and Viton-A showed deterioration following the first cycling at 400 torr. Using commercially available materials, the Tritium Systems Test Assembly (TSTA) designed and built a Portable Water Removal (PWR) Unit to reduce tritium oxide emissions during glovebox breaches. The PWR removes 99.9% of all tritium and saves between 0.7 and 3.5 curies of tritium oxide from being stacked during each of the five tests. Finally, a series of tests are done to determine whether the presence of SF{sub 6} changes the ability of palladium and platinum to catalyze the T{sub 2}-O{sub 2} reaction to form T{sub 2}O. No deterioration of the catalytic activity is observed. This is important because the Tokamak Fusion Test Reactor (TFTR) requires information about the effect of SF{sub 6}, an electrical insulator, on the catalytic behavior of Pt and Pd in a T{sub 2} environment. This information is necessary for the accident analysis in the Safety Analysis Report for TFTR. This study is done using an apparatus supplied to TSTA by TFTR.

  5. Byron Extended Facility

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

    Byron Extended Facility Map

  6. Ashton Extended Facility

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

    Ashton Extended Facility Map

  7. Tritium production from a plasma discharge on palladium

    SciTech Connect (OSTI)

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

    1995-12-01

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

  8. User Facilities

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

    User Facilities User Facilities User facility agreements allow Los Alamos partners and other entities to conduct research at our unique facilities. In 2011, LANL hosted more than 1,200 users at CINT, LANSCE, and NHMFL. Users came from across the DOE complex, from international academia, and from industrial companies from 45 states across the U.S. Unique world-class user facilities foster rich research opportunities Through its technology transfer efforts, LANL can implement user facility

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

    DOE Patents [OSTI]

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

    1982-08-19

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

  10. Evaluation of Tritium Content and Release from Pressurized Water Reactor Fuel Cladding

    SciTech Connect (OSTI)

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

    2015-09-01

    It is expected that tritium pretreatment will be required in future reprocessing plants to prevent the release of tritium to the environment (except for long-cooled fuels). To design and operate future reprocessing plants in a safe and environmentally compliant manner, the amount and form of tritium in the used nuclear fuel (UNF) must be understood and quantified. Tritium in light water reactor (LWR) fuel is dispersed between the fuel matrix and the fuel cladding, and some tritium may be in the plenum, probably as tritium labelled water (THO) or T2O. In a standard processing flowsheet, tritium management would be accomplished by treatment of liquid streams within the plant. Pretreating the fuel prior to dissolution to release the tritium into a single off-gas stream could simplify tritium management, so the removal of tritium in the liquid streams throughout the plant may not be required. The fraction of tritium remaining in the cladding may be reduced as a result of tritium pretreatment. Since Zircaloy® cladding makes up roughly 25% by mass of UNF in the United States, processes are being considered to reduce the volume of reprocessing waste for Zircaloy® clad fuel by recovering the zirconium from the cladding for reuse. These recycle processes could release the tritium in the cladding. For Zircaloy-clad fuels from light water reactors, the tritium produced from ternary fission and other sources is expected to be divided between the fuel, where it is generated, and the cladding. It has been previously documented that a fraction of the tritium produced in uranium oxide fuel from LWRs can migrate and become trapped in the cladding. Estimates of the percentage of tritium in the cladding typically range from 0–96%. There is relatively limited data on how the tritium content of the cladding varies with burnup and fuel history (temperature, power, etc.) and how pretreatment impacts its release. To gain a better understanding of how tritium in cladding

  11. Is Tritium Over-Regulated, Part 2 Should The TFG Support Higher...

    Office of Environmental Management (EM)

    Threshold Values? Presentation from the 33rd Tritium Focus Group Meeting held in Aiken, South Carolina on April 22-24, 2014. PDF icon Is Tritium Over-Regulated, Part 2...

  12. SEIS for the Production of Tritium in a Commercial Light Water...

    National Nuclear Security Administration (NNSA)

    SEIS for the Production of Tritium in a Commercial Light Water Reactor The NNSA, a ... the Production of Tritium in a Commercial Light Water Reactor (CLWR EIS; DOEEIS-0288). ...

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

    Broader source: Energy.gov [DOE]

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

  14. TRITIUM EFFECTS ON DYNAMIC MECHANICAL PROPERTIES OF POLYMERIC MATERIALS

    SciTech Connect (OSTI)

    Clark, E

    2008-11-12

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

  15. Alternate Tritium Production Methods Using A Liquid Lithium Target

    SciTech Connect (OSTI)

    Wilson, J.

    2015-10-08

    For over 60 years, the Savannah River Site’s primary mission has been the production of tritium. From the beginning, the Savannah River National Laboratory (SRNL) has provided the technical foundation to ensure the successful execution of this critical defense mission. SRNL has developed most of the processes used in the tritium mission and provides the research and development necessary to supply this critical component. This project was executed by first developing reactor models that could be used as a neutron source. In parallel to this development calculations were carried out testing the feasibility of accelerator technologies that could also be used for tritium production. Targets were designed with internal moderating material and optimized target was calculated to be capable of 3000 grams using a 1400 MWt sodium fast reactor, 850 grams using a 400 MWt sodium fast reactor, and 100 grams using a 62 MWt reactor, annually.

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

    Office of Energy Efficiency and Renewable Energy (EERE)

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

  17. Tritium on Z: The challenges and possibilities for MagLIF

    Broader source: Energy.gov [DOE]

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

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

    Broader source: Energy.gov [DOE]

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

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

    Broader source: Energy.gov [DOE]

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

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

    Broader source: Energy.gov [DOE]

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

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

    Broader source: Energy.gov [DOE]

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

  2. Tritium Aging Studies of LaNi4.15Al0.85 (LANA.85)

    Broader source: Energy.gov [DOE]

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

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

    Broader source: Energy.gov [DOE]

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

  4. Methods to Improve the Lower Limit of Detectionfor Tritium in the Air and on Surfaces

    Broader source: Energy.gov [DOE]

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

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

    Office of Energy Efficiency and Renewable Energy (EERE)

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

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

    Office of Energy Efficiency and Renewable Energy (EERE)

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

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

    Broader source: Energy.gov [DOE]

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

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

    Broader source: Energy.gov [DOE]

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

  9. Methods to Improve the Lower Limit of Detection for Tritium in the Air and on Surfaces

    Broader source: Energy.gov [DOE]

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

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

    Broader source: Energy.gov [DOE]

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

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

    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.

  12. Tritium Permeability of Incoloy 800H and Inconel 617

    SciTech Connect (OSTI)

    Philip Winston; Pattrick Calderoni; Paul Humrickhouse

    2011-09-01

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

  13. Design of the OMEGA Laser Target Chamber Tritium Removal System

    SciTech Connect (OSTI)

    Nobile, Arthur; Reichert, Heidi; Janezic, Roger T.; Harding, David R.; Lund, Lance D.; Shmayda, Walter T.

    2003-06-15

    Preparations are currently underway at the OMEGA laser at the University of Rochester Laboratory for Laser Energetics (UR/LLE) to conduct direct drive laser implosion campaigns with inertial confinement fusion targets containing deuterium-tritium (DT) cryogenic ice layers. The OMEGA Cryogenic Target Handling System will fill plastic targets with high-pressure DT (150 MPa) at 300 to 500 K, cool them down to cryogenic temperature (<25 K), form the DT ice layer, and transport the targets to the OMEGA laser target chamber. Targets will then be shot with the 60-beam 30-kJ OMEGA laser. A tritium removal system has been designed to remove tritium from effluents associated with operation of the target chamber and its associated diagnostic antechambers, vacuum pumping systems, and target insertion systems. The design of the target chamber tritium removal system (TCTRS) is based on catalytic oxidation of DT and tritiated methane to tritiated water (DTO), followed by immobilization of DTO on molecular sieves. The design of the TCTRS presented a challenge due to the low tritium release limits dictated by the tritium license at UR/LLE. Aspen Plus, a commercial software package intended for the simulation and design of chemical processing systems operating at steady state, was used to simulate and design the TCTRS. A second commercial software package, Aspen ADSIM, was used to simulate and design the TCTRS molecular sieve beds, which operate at unsteady state. In this paper, we describe the design of the TCTRS and the benefits that were realized by use of the Aspen Plus and Aspen ADSIM software packages.

  14. Tritium Permeability of Incoloy 800H and Inconel 617

    SciTech Connect (OSTI)

    Philip Winston; Pattrick Calderoni; Paul Humrickhouse

    2012-07-01

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

  15. 32nd Tritium Focus Group Meeting, Cloverleaf Building, Germantown MD, April 23-25, 2013

    Broader source: Energy.gov [DOE]

    The Tritium Focus Group (TFG), is a long standing DOE Working Group, whose purpose 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.

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

    SciTech Connect (OSTI)

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

    2008-07-15

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

  17. 36th Tritium Focus Group Meeting, Los Alamos, New Mexico, November 03-05, 2015

    Broader source: Energy.gov [DOE]

    The Tritium Focus Group (TFG), is a long standing DOE Working Group, whose purpose 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.

  18. 33rd Tritium Focus Group Meeting, Aiken, South Carolina, April 22-24, 2014

    Broader source: Energy.gov [DOE]

    The Tritium Focus Group (TFG), is a long standing DOE Working Group, whose purpose 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.

  19. User Facilities

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

    User Facilities User Facilities A new research frontier awaits! Our door is open, and we thrive on mutually beneficial partnerships and collaborations that drive innovations and new technologies. Unique world-class user facilities foster rich research opportunities Through its technology transfer efforts, Los Alamos National Laboratory can implement user facility agreements that allow its partners and other entities to conduct research at many of its unique facilities. While our largest user

  20. Facility Representatives

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

    2011-03-01

    This standard, DOE-STD-1063, Facility Representatives, defines the duties, responsibilities and qualifications for Department of Energy (DOE) Facility Representatives, based on facility hazard classification; risks to workers, the public, and the environment; and the operational activity level. This standard provides the guidance necessary to ensure that DOE’s hazardous nuclear and non-nuclear facilities have sufficient staffing of technically qualified facility representatives (FRs) to provide day-to-day oversight of contractor operations.

  1. 35th Tritium Focus Group Meeting, Princeton, New Jersey, May 05-07, 2015 |

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

    Department of Energy 5th Tritium Focus Group Meeting, Princeton, New Jersey, May 05-07, 2015 35th Tritium Focus Group Meeting, Princeton, New Jersey, May 05-07, 2015 35th Tritium Focus Group Meeting, Princeton, New Jersey, May 05-07, 2015 The Tritium Focus Group (TFG), is a long standing DOE Working Group, whose purpose is to promote cost-effective improvements in tritium safety, handling, transportation, storage, and operations, and to enhance communication across the Department of Energy

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

    Office of Environmental Management (EM)

    In-Situ Tritium Transport Parameters for Type 316 Stainless Steel during Irradiation D.J. Senor, W.G. Luscher K.K. Clayton, G.R. Longhurst Tritium Focus Group Meeting Savannah River National Laboratory Aiken, SC 23 April 2014 PNNL-SA-102143 Motivation and Scope TMIST-2 Experiment Measured in-reactor steady state tritium permeation through Type 316 stainless steel as a function of tritium partial pressure and temperature Tritium permeation irradiation enhancement of ~3X was observed relative to

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

    SciTech Connect (OSTI)

    Powell, G.L.; Laesser, R.

    1988-12-20

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

  4. Preparation and characterization of tritium targets. [Titanium tritide

    SciTech Connect (OSTI)

    Ramey, D.W.; Adair, H.L.

    1982-01-01

    The Isotope Research Materials Laboratory (IRML) of the Oak Ridge National Laboratory (ORNL) prepares tritium targets that are used to produce an intense beam of 14.5 MeV neutrons by the /sup 3/H(/sup 2/H,/sup 1/n)/sup 4/He reaction. The intense beams of 14.5 MeV neutrons are used in programs involving cancer research, materials evaluation, and materials identification. The tritium targets required by researchers have ranged in size from 1-cm diam to 50-cm diam and have contained from 3.7 x 10/sup 4/ MBq to 2.2 x 10/sup 8/ MBq of tritium. Important parameters in the performance of tritium targets, when bombarded with deuterons, include target lifetime and neutron output. These parameters are heavily dependent on the host metal layer used to form the tritide compound and the gas-to-metal loading of the host material. Fabrication procedures used in preparing titanium tritide targets with subsequent gas-to-metal determinations are described.

  5. Savannah River Site by the Numbers August 2015

    Office of Environmental Management (EM)

    Also built were a number of support facilities including two chemical separations plants, a heavy water extraction plant, a nuclear fuel and target fabrication facility, a tritium ...

  6. Plutonium Uranium Extraction Plant (PUREX) - Hanford Site

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

    About Us Projects & Facilities Plutonium Uranium Extraction Plant (PUREX) About Us About ... and 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage ...

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

    SciTech Connect (OSTI)

    Winterberg, F.

    2009-01-01

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

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

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

    Winterberg, F.

    2009-01-01

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

  9. ORISE: Facilities

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

    ORISE Facilities Unique laboratories and training centers among the assets managed on behalf of the U.S. Department of Energy The Oak Ridge Institute for Science and Education (ORISE) is home to a number of on- and off-site facilities that support the U.S. Department of Energy's (DOE) science education and research mission. From on-site medical laboratories to radiation emergency medicine training facilities, ORISE facilities are helping to address national needs in the following areas:

  10. Science Facilities

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

    Facilities Science Facilities The focal point for basic and applied R&D programs with a primary focus on energy but also encompassing medical, biotechnology, high-energy physics, and advanced scientific computing programs. Center for Integrated Nanotechnologies» Dual Axis Radiographic Hydrodynamic Test Facility (DARHT)» Electron Microscopy Lab» Ion Beam Materials Lab» Isotope Production Facility» Los Alamos Neutron Science Center» Lujan Center» Matter-Radiation Interactions in

  11. Facility Safety

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

    1996-10-24

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

  12. Facility Safety

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

    1995-11-16

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

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

    SciTech Connect (OSTI)

    Chang H. Oh; Eung S. Kim

    2011-03-01

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

  14. Retention and release of tritium in aluminum clad, Al-Li alloys

    SciTech Connect (OSTI)

    Louthan, M.R. Jr.

    1991-12-31

    Tritium retention in and release from aluminum clad, aluminum-lithium alloys is modeled from experimental and operational data developed during the thirty plus years of tritium production at the Savannah River Site. The model assumes that tritium atoms, formed by the {sup 6}Li(n,{alpha}){sup 3}He reaction, are produced in solid solution in the Al-Li alloy. Because of the low solubility of hydrogen isotopes in aluminum alloys, the irradiated Al-Li rapidly becomes supersaturated in tritium. Newly produced tritium atoms are trapped by lithium atoms to form a lithium tritide. The effective tritium pressure required for trap or tritide stability is the equilibrium decomposition pressure of tritium over a lithium tritide-aluminum mixture. The temperature dependence of tritium release is determined by the permeability of the cladding to tritium and the local equilibrium at the trap sites. This model is used to calculate tritium release from aluminum clad, aluminum-lithium alloys. 9 refs., 3 figs.

  15. Retention and release of tritium in aluminum clad, Al-Li alloys

    SciTech Connect (OSTI)

    Louthan, M.R. Jr.

    1991-01-01

    Tritium retention in and release from aluminum clad, aluminum-lithium alloys is modeled from experimental and operational data developed during the thirty plus years of tritium production at the Savannah River Site. The model assumes that tritium atoms, formed by the {sup 6}Li(n,{alpha}){sup 3}He reaction, are produced in solid solution in the Al-Li alloy. Because of the low solubility of hydrogen isotopes in aluminum alloys, the irradiated Al-Li rapidly becomes supersaturated in tritium. Newly produced tritium atoms are trapped by lithium atoms to form a lithium tritide. The effective tritium pressure required for trap or tritide stability is the equilibrium decomposition pressure of tritium over a lithium tritide-aluminum mixture. The temperature dependence of tritium release is determined by the permeability of the cladding to tritium and the local equilibrium at the trap sites. This model is used to calculate tritium release from aluminum clad, aluminum-lithium alloys. 9 refs., 3 figs.

  16. Behaviour of tritium in the vacuum vessel of JT-60U

    SciTech Connect (OSTI)

    Kobayashi, K.; Miya, N.; Ikeda, Y.; Torikai, Y.; Saito, M.; Alimov, V.

    2015-03-15

    The disassembly of the JT-60U torus started in 2010 after 18 years of deuterium plasma operations. The vessel is made of Inconel 625. Therefore, it was very important to study the hydrogen isotope (particularly tritium) behavior in Inconel 625 from the viewpoint of the clearance procedure. Inconel 625 specimen was exposed to the D{sub 2} (92.8 %) - T{sub 2} (7.2 %) gas mixture at 573 K for 5 hours. The tritium release from the specimen at 298 K was controlled for about 1 year. After that a part of tritium remaining in the specimen was released by heating up to 1073 K. Other part of tritium trapped in the specimen was measured by chemical etching method. Most of the chemical form of the released tritium was HTO. The contaminated specimen by tritium was released continuously the diffusible tritium under the ambient condition. In the tritium release experiment, the amount of desorbed tritium was about 99% during 1 year. It was considered that the tritium in Inconel 625 was released easily.

  17. Wheelabrator Millbury Facility Biomass Facility | Open Energy...

    Open Energy Info (EERE)

    Facility Facility Wheelabrator Millbury Facility Sector Biomass Facility Type Municipal Solid Waste Location Worcester County, Massachusetts Coordinates 42.4096528, -71.8571331...

  18. 3HE RECOVERY FROM A TRITIUM-AGED LANA75 SAMPLE

    SciTech Connect (OSTI)

    Shanahan, K.

    2010-12-01

    {sup 3}He recovery is a topic of recent interest. One potential recovery source is from metal hydride materials once used to store tritium, as the decay product, {sup 3}He, is primarily trapped in the metal lattice, usually in bubbles, with such materials. In 2001, a Tritium Exposure Program (TEP) sample known as LANA75-SP1 was retired and the material was removed from the test cell and stored. Subsequently scoping temperature programmed desorption (TPD) experiments were conducted on that material to see what it might take to drive out He and residual H isotopes (the heel). Two experiments consisted of heating the sample in the presence of an excess of tin (the so-called Sn fusion experiment), and one was a simple TPD with no additives. Prior data on the so-called '21-month bed' material in the 1980's had produced {approx}21 cc of gas per gram of a LANA30 material (LaNi4.7Al0.3), with approximately 67% of that being {sup 3}He and the rest being D{sub 2} (Fig.3). However, the material had to be heated in excess of 850 C to obtain that level. Heating to less produced approximately half that amount of gas. The data also showed that {sup 3}He was released at different temperatures than the residual hydrogen isotopes. Unfortunately this implies full {sup 3}He recovery will be a difficult process. Therefore, it seemed advisable to attempt to extract as much information from the 3 scoping experiments from 2001-2 as possible.

  19. Elastic Electron Scattering from Tritium and Helium-3

    DOE R&D Accomplishments [OSTI]

    Collard, H.; Hofstadter, R.; Hughes, E. B.; Johansson, A.; Yearian, M. R.; Day, R. B.; Wagner, R. T.

    1964-10-01

    The mirror nuclei of tritium and helium-3 have been studied by the method of elastic electron scattering. Absolute cross sections have been measured for incident electron energies in the range 110 - 690 MeV at scattering angles lying between 40 degrees and 135 degrees in this energy range. The data have been interpreted in a straightforward manner and form factors are given for the distributions of charge and magnetic moment in the two nuclei over a range of four-momentum transfer squared 1.0 - 8.0 F{sup -2}. Model-independent radii of the charge and magnetic moment distributions are given and an attempt is made to deduce form factors describing the spatial distribution of the protons in tritium and helium-3.

  20. FINITE ELEMENT ANALYSIS OF BULK TRITIUM SHIPPING PACKAGE

    SciTech Connect (OSTI)

    Jordan, J.

    2010-06-02

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

  1. Tritium calibration of the LUX dark matter experiment

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

    Akerib, D. S.

    2016-04-20

    Here, we present measurements of the electron-recoil (ER) response of the LUX dark matter detector based upon 170,000 highly pure and spatially uniform tritium decays. We reconstruct the tritium energy spectrum using the combined energy model and find good agreement with expectations. We report the average charge and light yields of ER events in liquid xenon at 180 and 105 V/cm and compare the results to the NEST model. We also measure the mean charge recombination fraction and its fluctuations, and we investigate the location and width of the LUX ER band. These results provide input to a reanalysis ofmore » the LUX run 3 weakly interacting massive particle search.« less

  2. DEVELOPMENT AND USE OF A BULK TRITIUM SHIPPING PACKAGE

    SciTech Connect (OSTI)

    Blanton, P.

    2010-09-30

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

  3. ARM - Facility News Article

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

    July 15, 2008 [Facility News] Extract This! Enhanced Visualization Tool Available at the Data Archive Bookmark and Share Custom 2-day plot of downwelling shortwave and longwave radiation made using the enhanced NCVweb feature. Like many scientific organizations, the ARM Data Archive stores and distributes atmospheric data from the ARM sites in Network common data form, or NetCDF. This file format applies names or attributes to the various layers of data for efficient identification and

  4. APPLICATION OF RISK MANAGEMENT PRACTICES TO NNSA TRITIUM READINESS SUBPROGRAM

    SciTech Connect (OSTI)

    Shete, S; Srini Venkatesh, S

    2007-01-31

    The National Nuclear Security Administration (NNSA), Office of Stockpile Technology (NNSA/NA-123) chartered a risk assessment of the Tritium Readiness (TR) Subprogram to identify risks and to develop handling strategies with specific action items that could be scheduled and tracked to completion in order to minimize program failures. This assessment was performed by a team of subject matter experts (SMEs) comprised of representatives from various organizations participating in the TR Subprogram. The process was coordinated by Savannah River Site, Systems Engineering (SRS/SE) with support from Subprogram Team. The Risk Management Process steps performed during this risk assessment were: Planning, Identification, Grading, Handling, and Impact Determination. All of the information captured during the risk assessment was recorded in a database. The team provided estimates for the cost and schedule impacts of implementing the recommended handling strategies and facilitated the risk based cost contingency analysis. The application of the Risk Management Practices to the NNSA Tritium Readiness Subprogram resulted in: (1) The quarterly review and update of the Risk Management Database to include an evaluation of all existing risks and the identification/evaluation of any potential new risks. (2) The risk status and handling strategy action item tracking mechanism that has visibility and buy-in throughout the Tritium Readiness Subprogram to ensure that approved actions are completed as scheduled and that risk reduction is being achieved. (3) The generation of a risk-based cost contingency estimate that may be used by the Tritium Readiness Subprogram Manager in establishing future year program budgets.

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

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

    Luis Alvarez, the Hydrogen Bubble Chamber, Tritium, and Dinosaurs Resources with Additional Information * Patents Luis Alvarez Courtesy Lawrence Berkeley National Laboratory 'Luis W. Alvarez was an adventurer physicist. The two terms may seem an odd combination until one considers Alvarez's career. A member of the National Inventor's Hall of Fame, Alvarez developed the proton linear accelerator, patented three types of radar still used today, designed an instrument that for 15 years served as

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

    SciTech Connect (OSTI)

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

    2013-10-02

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

  7. Tritium and neutron measurements of a solid state cell

    SciTech Connect (OSTI)

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

    1989-01-01

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

  8. Measurement of uptake and release of tritium by tungsten

    SciTech Connect (OSTI)

    Nakayama, M.; Torikai, Y.; Saito, M.; Penzhorn, R.D.; Isobe, K.; Yamanishi, T.; Kurishita, H.

    2015-03-15

    Tungsten is currently contemplated as plasma facing material for the divertor of future fusion machines. In this paper the uptake of tritium by tungsten and its release behavior have been investigated. Tungsten samples have been annealed at various temperatures and loaded at also different temperatures with deuterium containing 7.2 % tritium at a pressure of 1.2 kPa. A specific system was designed to assess the release of tritiated water and molecular tritium by the samples. Due to the rather low solubility of hydrogen isotopes in tungsten it is particularly important to be aware of the presence of hydrogen traps or thin oxide films. As shown in this work, traps or oxide films may affect the retention capability of tungsten and lead to significantly modified release properties. It became clear that there were capture sites that had different thermal stability and different capture intensity in tungsten after polishing, or oxide films that were grown on the surface of tungsten and had barrier effects.

  9. Tritium processing for the European test blanket systems: current status of the design and development strategy

    SciTech Connect (OSTI)

    Ricapito, I.; Calderoni, P.; Poitevin, Y.; Aiello, A.; Utili, M.; Demange, D.

    2015-03-15

    Tritium processing technologies of the two European Test Blanket Systems (TBS), HCLL (Helium Cooled Lithium Lead) and HCPB (Helium Cooled Pebble Bed), play an essential role in meeting the main objectives of the TBS experimental campaign in ITER. The compliancy with the ITER interface requirements, in terms of space availability, service fluids, limits on tritium release, constraints on maintenance, is driving the design of the TBS tritium processing systems. Other requirements come from the characteristics of the relevant test blanket module and the scientific programme that has to be developed and implemented. This paper identifies the main requirements for the design of the TBS tritium systems and equipment and, at the same time, provides an updated overview on the current design status, mainly focusing onto the tritium extractor from Pb-16Li and TBS tritium accountancy. Considerations are also given on the possible extrapolation to DEMO breeding blanket. (authors)

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

    SciTech Connect (OSTI)

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

    2013-09-01

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

  11. Method and device for secure, high-density tritium bonded with carbon

    DOE Patents [OSTI]

    Wertsching, Alan Kevin; Trantor, Troy Joseph; Ebner, Matthias Anthony; Norby, Brad Curtis

    2016-04-05

    A method and device for producing secure, high-density tritium bonded with carbon. A substrate comprising carbon is provided. A precursor is intercalated between carbon in the substrate. The precursor intercalated in the substrate is irradiated until at least a portion of the precursor, preferably a majority of the precursor, is transmutated into tritium and bonds with carbon of the substrate forming bonded tritium. The resulting bonded tritium, tritium bonded with carbon, produces electrons via beta decay. The substrate is preferably a substrate from the list of substrates consisting of highly-ordered pyrolytic graphite, carbon fibers, carbon nanotunes, buckministerfullerenes, and combinations thereof. The precursor is preferably boron-10, more preferably lithium-6. Preferably, thermal neutrons are used to irradiate the precursor. The resulting bonded tritium is preferably used to generate electricity either directly or indirectly.

  12. In-bed accountability of tritium in production scale metal hydride storage beds

    SciTech Connect (OSTI)

    Klein, J.E.

    1995-10-01

    An `in-bed accountability` (IBA) flowing gas calorimetric measurement method has been developed and implemented to eliminate the need to remove tritium from production scale metal hydride storage beds for inventory measurement purposes. Six-point tritium IBA calibration curves have been completed for two, 390 gram tritium metal hydride storage beds. The calibration curves for the two tritium beds are similar to those obtained from the `cold` test program. Tritium inventory errors at the 95 percent confidence level ranged from {+-} 7.3 to 8.6 grams for the cold test results compared to {+-} 4.2 to 7.5 grams obtained for the two tritium calibrated beds. 5 refs., 4 figs., 1 tab.

  13. Tritium permeation characterization of materials for fusion and generation IV very high temperature reactors

    SciTech Connect (OSTI)

    Thomson, S.; Pilatzke, K.; McCrimmon, K.; Castillo, I.; Suppiah, S.

    2015-03-15

    The objective of this work is to establish the tritium-permeation properties of structural alloys considered for Fusion systems and very high temperature reactors (VHTR). A description of the work performed to set up an apparatus to measure permeation rates of hydrogen and tritium in 304L stainless steel is presented. Following successful commissioning with hydrogen, the test apparatus was commissioned with tritium. Commissioning tests with tritium suggest the need for a reduction step that is capable of removing the oxide layer from the test sample surfaces before accurate tritium-permeation data can be obtained. Work is also on-going to clearly establish the temperature profile of the sample to correctly estimate the tritium-permeability data.

  14. Beamlines & Facilities

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

    Imaging Group: Beamlines The X-ray Micrscopy and Imaging Group operates several beamlines and facilities. The bending magnet beamline (2-BM) entertaines 2 general user programs in...

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

    SciTech Connect (OSTI)

    1997-03-01

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

  16. Radioluminescent light sources, tritium containing polymers, and methods for producing the same

    DOE Patents [OSTI]

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

    1989-12-26

    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.

  17. Radioluminescent light sources, tritium containing polymers, and methods for producing the same

    DOE Patents [OSTI]

    Jensen, George A. (Richland, WA); Nelson, David A. (Richland, WA); Molton, Peter M. (Richland, WA)

    1989-01-01

    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 matix by a tritium containing polymer. The tritium containing polymer comprises a polymer backbone which is essentially void of tritium.

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

    DOE Patents [OSTI]

    Meyer, Thomas J.; Narula, Poonam M.

    2001-01-01

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

  19. Tritium Laboratory Karlsruhe: administrative and technical framework for isotope laboratory operation

    SciTech Connect (OSTI)

    Welte, S.; Besserer, U.; Osenberg, D.; Wendel, J.

    2015-03-15

    Originally licensed in 1993 the Tritium Laboratory Karlsruhe (TLK) is a unique pilot scale isotope laboratory focused on tritium handling and processing to conduct a variety of scientific experiments and development tasks in view of future fusion power plants. TLK currently operates 15 glove boxes of 125 m{sup 3} total volume in an experimental hall measuring nearly 1500 m{sup 2}. The tritium infrastructure, comprising of the tritium storage system, the tritium transfer system and the isotope separation system, is integrated into TLK as a closed loop system to supply tritium to the experiments. Having a license for handling of up to 40 g of tritium and a closed tritium processing loop, TLK is a unique institute in non-military tritium research. In order to fulfil all requirements regarding the license, a framework of regulations is applied as a basis for the operation of TLK, as well as the setup of new experiments and the design of components. This paper will give an overview on the framework of operation in view of licensing issues, as well as administrative and technical regulations mandatory to legally and reliably operate an isotope laboratory of this scale.

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

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

    in a Commercial Light Water Reactor Supplemental Environmental Impact Statement EIS-0288-S1: Production of Tritium in a Commercial Light Water Reactor Supplemental Environmental ...

  1. Facility Representatives

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

    2006-04-06

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

  2. Facility Safety

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

    2005-12-22

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

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

    SciTech Connect (OSTI)

    Hiergesell, Robert A.

    2008-01-15

    trans-river flow is eastward or westward depends primarily on the position of the Savannah River as it meanders back and forth within the flood plain and is limited to narrow sections of land adjacent to the river. With respect to the only location of westward trans-river flow that has a recharge area within the SRS, the new evaluations of hypothetical pumping scenarios indicated that only a very slight impact is incurred, even under the most extreme groundwater extraction scenario. The updated model did not result in a significant change in the location of the recharge areas at SRS and the only impact was measured in slight changes in the travel times associated with the travel path. The median groundwater travel times for particles released under each of the 4 groundwater extraction scenarios ranged from 366 to 507 years while. Under the most extreme scenario, that under which SRS groundwater extraction is discontinued, the shortest travel time was reduced from 90 to 79 years. It should be emphasized that the groundwater transit times do not include the time required for groundwater to migrate vertically downward across the uppermost aquifer (i.e. at the recharge area), thus the actual groundwater travel times could be up to several decades longer than what was calculated in the model. The exhaustive evaluations that have been conducted indicates that it is highly unlikely that tritiated groundwater originating at the SRS could migrate into Georgia and explain the low tritium activity levels that were originally observed in certain domestic water supply wells. Considering that those wells were located at some distance (several km) from the Savannah River, a far more likely explanation is that tritiated rainfall infiltrated the subsurface and recharged the shallow aquifer within which the well was finished.

  4. Facility Safety

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

    2012-12-04

    The Order establishes facility and programmatic safety requirements for DOE and NNSA for nuclear safety design criteria, fire protection, criticality safety, natural phenomena hazards (NPH) mitigation, and System Engineer Program. This Page Change is limited in scope to changes necessary to invoke DOE-STD-1104, Review and Approval of Nuclear Facility Safety Basis and Safety Design Basis Document, and revised DOE-STD-3009-2014, Preparation of Nonreactor Nuclear Facility Documented Safety Analysis as required methods. DOE O 420.1C Chg 1, dated 2-27-15, supersedes DOE O 420.1C.

  5. Is Tritium Over-regulated, Part 2: Should the TFG Support Higher Tritium Threshold Values?

    Office of Environmental Management (EM)

    Investing in a New Era of Manufacturing Technology Investing in a New Era of Manufacturing Technology June 24, 2011 - 6:05pm Addthis John Schueler John Schueler Former New Media Specialist, Office of Public Affairs What does this mean for me? The Energy Department will be investing up to $120 million over three years in the development of transformational manufacturing technologies and innovative materials that could enable industrial facilities to dramatically increase their energy efficiency.

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

    SciTech Connect (OSTI)

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

    2005-08-30

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

  7. Gas Utilization Facility Biomass Facility | Open Energy Information

    Open Energy Info (EERE)

    Gas Utilization Facility Biomass Facility Jump to: navigation, search Name Gas Utilization Facility Biomass Facility Facility Gas Utilization Facility Sector Biomass Facility Type...

  8. Total Energy Facilities Biomass Facility | Open Energy Information

    Open Energy Info (EERE)

    Energy Facilities Biomass Facility Jump to: navigation, search Name Total Energy Facilities Biomass Facility Facility Total Energy Facilities Sector Biomass Facility Type...

  9. Facility Safety

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

    2002-05-20

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

  10. Facility Safety

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

    2005-12-22

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

  11. Facility Safety

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

    2013-06-21

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

  12. Facility Safety

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

    2000-11-20

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

  13. EFFECTS OF ONE WEEK TRITIUM EXPOSURE ON EPDM ELASTOMER

    SciTech Connect (OSTI)

    Clark, E

    2007-06-07

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

  14. Field Deployable Tritium Assay System Host Graphical User Interface Software

    Energy Science and Technology Software Center (OSTI)

    1998-05-12

    The FDTASHOST software is a Graphical User Interface for the Field Deployable Tritium Assay System (FDTAS - Invention Disclosure SRS-96-09-091 has been submitted). The program runs on the Host computer which is located in the Laboratory and connected to the FDTAS remote field system via a modem over a phone line. The operator receives status information and messages from the Remote system. The operator can enter in commands to be executed by the remote systemmore » using the mouse and a pull down menu.« less

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

    Broader source: Energy.gov [DOE]

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

  16. Working with SRNL - Our Facilities - Glovebox Facilities

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

    SRNL Our Facilities - Glovebox Facilities Govebox Facilities are sealed, protectively-lined compartments with attached gloves, allowing workers to safely handle dangerous materials...

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

    SciTech Connect (OSTI)

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

    1997-04-01

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

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

    DOE Patents [OSTI]

    Cawley, William E.; Trapp, Turner J.

    1984-10-09

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

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

    DOE Patents [OSTI]

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

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

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

    DOE Patents [OSTI]

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

    1984-10-09

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

  1. SLAC Accelerator Test Facilities

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

    FACET & TF Careers & Education Archived FACET User Facility Quick Launch About FACET & Test Facilities Expand About FACET & Test Facilities FACET & Test Facilities User Portal...

  2. EIS-0288-S1: Production of Tritium in a Commercial Light Water Reactor Supplemental Environmental Impact Statement

    Broader source: Energy.gov [DOE]

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

  3. Design and tritium permeation analysis of China HCCB TBM port cell

    SciTech Connect (OSTI)

    Jiangfeng, S.; Guoqiang, H.; Zhiyong, H.; Chang'an, C.; Deli, L.

    2015-03-15

    China is planning to develop a helium-cooled ceramic breeder (HCCB) test blanket module (TBM) on ITER to test key blanket technologies. In this paper, the design and tritium permeation analysis of China HCCB TBM port cell are introduced. A theoretical model has been developed to estimate tritium permeation rates and leak rates from the components and pipes which China has scheduled to house in the port cell. It is shown that on normal working conditions, the permeation and leak rate of the systems in the port cell will be no higher than 1.58 Ci/d without the use of tritium permeation barriers, and 0.10 Ci/d with the use of tritium permeation barriers. It also appears that tritium permeation barriers are necessary for high temperature components such as the reduction bed and the heater.

  4. Post service examination of turbomolecular pumps after stress testing with Kg-scale tritium throughput

    SciTech Connect (OSTI)

    Priester, F.; Roelling, M.

    2015-03-15

    Turbomolecular pumps (TMP) will be used with large amounts of tritium in future fusion machines like ITER, DEMO and in the KATRIN Experiment. In this work, a stress test of a large, magnetically levitated TMP (Leybold MAG W2800) with a tritium throughput of 1.1 kg over 384 days of operation was performed at TLK. After this, the pump was dismantled and the tritium uptake in several parts was deter-mined. Especially the non-metallic parts of the pump have absorbed large amounts of tritium and are most likely responsible for the observed pollution of the process gas. The total tritium uptake of the TMP was estimated with 0.1-1.1 TBq. No radiation-induced damages were found on the inner parts of the pump. The TMP showed no signs of functional limitations during the 384 days of operation. (authors)

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

    SciTech Connect (OSTI)

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

    1995-01-01

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

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

    SciTech Connect (OSTI)

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

    1995-10-01

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

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

    SciTech Connect (OSTI)

    Ide, T.; Okuno, K.; Konishi, S.; Sakai, F.; Fukui, H.; Enoeda, M.; Naruse, Y.; Anderson, J.L.; Bartlit, J.R.; Los Alamos National Lab., NM )

    1989-01-01

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

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

    SciTech Connect (OSTI)

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

    2005-07-15

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

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

    SciTech Connect (OSTI)

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

    2004-01-01

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

  10. Weldability of tritium-charged 304L stainless steel

    SciTech Connect (OSTI)

    Kanne, W R; Angerman, C L; Eberhard, B J

    1987-02-01

    Attempts to repair the wall of C-Reactor Tank at the Savannah River Plant were halted when Gas Tungsten Arc (GTA) welds joining patches to the wall developed toe cracks in the heat affected zone (HAZ). The cause of the toe cracks was investigated by welding on 304L samples that were tritium charged and aged to produce helium. Helium embrittlement was shown to be the likely cause of weld toe cracking in C-Reactor Tank. GTA welds made on helium impregnated 304L produced toe cracks identical to those that caused leaking patches during C-Reactor Tank repair. Heating of a sample to remove deuterium and tritium without removing helium did not reduce cracking susceptibility. Low heat input and spot GTA welds also produced cracks, indicating possible problems using these techniques for reactor repair. However, cracks were not produced by solid state resistance welds, or by a very low heat GTA pass that did not produce melting. This indicates that non-melting or low tensile stress techniques could be used for repair.

  11. Tritium-compatible high-vacuum pumping system

    SciTech Connect (OSTI)

    Coffin, D.O.

    1982-04-01

    Large magnetic-fusion power reactors will require vacuum pumps with greater speeds and capacities for hydrogen and helium isotopes than any now available. Present fusion engineering efforts concentrate on DT-burning tokamaks, so pumps for these systems must perform reliably during prolonged exposure to radioactive tritium gas. A mechanical pumping train that exposes only tritium-compatible metals to the vacuum medium has been assembled. The system comprises a reciprocating metal-bellows compressor, a bellows-sealed, spiral-cavity, positive-displacement blower, and a canned-motor, magnetic-bearing turbopump. These pumps feature hermetic drives, and their pumping chambers contain no liquid metals, plastics, elastomers, or lubricants. The pumping system achieves vacuums as low as 1 mPa without cryogenics, and its pumping speed increases from 5 L/s at atmospheric pressure to 500 L/s below 100 mPa. Systems with ultimate pumping speeds of 5000 L/s can be assembled using larger pumps of the same design as those tested.

  12. Tritium Specific Adsorption Simulation Utilizing the OSPREY Model

    SciTech Connect (OSTI)

    Veronica Rutledge; Lawrence Tavlarides; Ronghong Lin; Austin Ladshaw

    2013-09-01

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

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

    SciTech Connect (OSTI)

    Okhuysen, Brett S.; Pulliam, Elias Noel

    2015-11-05

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

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

    SciTech Connect (OSTI)

    WJ Apley

    1997-12-01

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

  15. Monte Carlo prompt dose calculations for the National Ingition Facility

    SciTech Connect (OSTI)

    Latkowski, J.F.; Phillips, T.W.

    1997-01-01

    During peak operation, the National Ignition Facility (NIF) will conduct as many as 600 experiments per year and attain deuterium- tritium fusion yields as high as 1200 MJ/yr. The radiation effective dose equivalent (EDE) to workers is limited to an average of 03 mSv/yr (30 mrem/yr) in occupied areas of the facility. Laboratory personnel determined located outside the facility will receive EDEs <= 0.5 mSv/yr (<= 50 mrem/yr). The total annual occupational EDE for the facility will be maintained at <= 0.1 person-Sv/yr (<= 10 person- rem/yr). To ensure that prompt EDEs meet these limits, three- dimensional Monte Carlo calculations have been completed.

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

    DOE Patents [OSTI]

    Heung, L.K.

    1994-03-29

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

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

    DOE Patents [OSTI]

    Heung, Leung K.

    1994-01-01

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

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

    SciTech Connect (OSTI)

    Chang H. Oh; Eung S. Kim

    2009-09-01

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

  19. Nuclear Facilities

    Broader source: Energy.gov [DOE]

    The nuclear sites list and map shows how DOE nuclear operations are mostly divided between nuclear weapons stockpile maintenance, research and environmental cleanup. The operations are performed within several different facilities supporting nuclear reactor operations, nuclear research, weapons disassembly, maintenance and testing, hot cell operations, nuclear material storage and processing and waste disposal.

  20. Facility Safety

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

    2012-12-04

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

  1. Facility Safety

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

    1995-10-13

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

  2. Beam Test Facility

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

    Beam Test Facility Beam Test Facility Print Tuesday, 20 October 2009 09:36 Coming Soon

  3. Mass Spectrometer Facility | Photosynthetic Antenna Research Center

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

    Mass Spectrometer Facility Mass Spectrometer Facility The PARC Mass Spectrometer Facility uses customized instrumentation to directly measure the individual polypeptide mass of different light-harvesting complexes to do assignment to specific gene products and investigate protein processing. Newly developed techniques are also applied to measure the mass of native protein complexes. Structural information of complexes is extracted by combining protein chemical modification and H/D exchange

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

    SciTech Connect (OSTI)

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

    2012-04-03

    The Defense Nuclear Facilities Safety Board (DNFSB) has recently questioned the appropriate value for tritium deposition velocity used in the MELCOR Accident Consequence Code System Ver. 2 (Chanin and Young 1998) code when estimating bounding dose (95th percentile) for safety analysis (DNFSB 2011). The purpose of this paper is to provide appropriate, defensible values of the tritium deposition velocity for use in Savannah River Site (SRS) safety analyses. To accomplish this, consideration must be given to the re-emission of tritium after deposition. Approximately 85% of the surface area of the SRS is forested. The majority of the forests are pine plantations, 68%. The remaining forest area is 6% mixed pine and hardwood and 26% swamp hardwood. Most of the path from potential release points to the site boundary is through forested land. A search of published studies indicate daylight, tritiated water (HTO) vapor deposition velocities in forest vegetation can range from 0.07 to 2.8 cm/s. Analysis of the results of studies done on an SRS pine plantation and climatological data from the SRS meteorological network indicate that the average deposition velocity during daylight periods is around 0.42 cm/s. The minimum deposition velocity was determined to be about 0.1 cm/s, which is the recommended bounding value. Deposition velocity and residence time (half-life) of HTO in vegetation are related by the leaf area and leaf water volume in the forest. For the characteristics of the pine plantation at SRS the residence time corresponding to the average, daylight deposition velocity is 0.4 hours. The residence time corresponding to the night-time deposition velocity of 0.1 cm/s is around 2 hours. A simple dispersion model which accounts for deposition and re-emission of HTO vapor was used to evaluate the impact on exposure to the maximally exposed offsite individual (MOI) at the SRS boundary (Viner 2012). Under conditions that produce the bounding, 95th percentile MOI exposure

  5. Metro Methane Recovery Facility Biomass Facility | Open Energy...

    Open Energy Info (EERE)

    Methane Recovery Facility Biomass Facility Jump to: navigation, search Name Metro Methane Recovery Facility Biomass Facility Facility Metro Methane Recovery Facility Sector Biomass...

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

    SciTech Connect (OSTI)

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

    2014-09-01

    To design and operate future reprocessing plants in a safe and environmentally compliant manner, the amount and form of tritium in the used nuclear fuel (UNF) must be understood and quantified.To gain a better understanding of how tritium in cladding will behave during processing, scoping tests are being performed to determine the tritium content in the cladding pre- and post-tritium pretreatment. A sample of Surry-2 pressurized water reactor (PWR) cladding was heated to 11001200C to oxidize the zirconium and release all of the tritium in the cladding sample. The tritium content was measured to be ~240 Ci/g. Cladding samples were heated to 500C, which is within the temperature range (480 - 600C) expected for standard air tritium pretreatment systems, and to a slightly higher temperature (700C) to determine the impact of tritium pretreatment on tritium release from the cladding. Heating at 500C for 24 hr removes ~0.2% of the tritium from the cladding, and heating at 700C for 24 hr removes ~9%. Thus, a significant fraction of the tritium remains bound in the cladding and must be considered in operations involving cladding recycle.

  7. Dependence of Tritium Release from Stainless Steel on Temperature and Water Vapor

    SciTech Connect (OSTI)

    Shmayda, W. T.; Sharpe, M.; Boyce, A. M.; Shea, R.; Petroski, B.; Schroeder, W. U.

    2015-09-15

    The impact of water vapor and temperature on the release of tritium from stainless steel was studied. Degreased stainless steel samples loaded with tritium at room temperature following a 24-h degassing in vacuum at room temperature were subjected to increasing temperatures or humidity. In general, increasing either the sample temperature or the humidity causes an increased quantity of tritium to be removed. Increasing the temperature to 300°C in a dry gas stream results in a significant release of tritium and is therefore an effective means for reducing the tritium inventory in steel. For humid purges at 30°C, a sixfold increase in humidity results in a tenfold increase in the peak outgassing rate. Increasing the humidity from 4 parts per million (ppm) to 1000 ppm when the sample temperature is 100°C causes a significant increase in the tritium outgassing rate. Finally, a simple calculation shows that only 15% of the activity present in the sample was removed in these experiments, suggesting that the surface layer of adsorbed water participates in regulating tritium desorption from the surface.

  8. Dependence of Tritium Release from Stainless Steel on Temperature and Water Vapor

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

    Shmayda, W. T.; Sharpe, M.; Boyce, A. M.; Shea, R.; Petroski, B.; Schroeder, W. U.

    2015-09-15

    The impact of water vapor and temperature on the release of tritium from stainless steel was studied. Degreased stainless steel samples loaded with tritium at room temperature following a 24-h degassing in vacuum at room temperature were subjected to increasing temperatures or humidity. In general, increasing either the sample temperature or the humidity causes an increased quantity of tritium to be removed. Increasing the temperature to 300°C in a dry gas stream results in a significant release of tritium and is therefore an effective means for reducing the tritium inventory in steel. For humid purges at 30°C, a sixfold increasemore » in humidity results in a tenfold increase in the peak outgassing rate. Increasing the humidity from 4 parts per million (ppm) to 1000 ppm when the sample temperature is 100°C causes a significant increase in the tritium outgassing rate. Finally, a simple calculation shows that only 15% of the activity present in the sample was removed in these experiments, suggesting that the surface layer of adsorbed water participates in regulating tritium desorption from the surface.« less

  9. Letter Report for Analytical Results for five Swipe Samples from the Northern Biomedical Research Facility, Muskegon Michigan

    SciTech Connect (OSTI)

    Ivey, Wade

    2013-12-17

    Oak Ridge Associated Universities (ORAU), under the Oak Ridge Institute for Science and Education (ORISE) contract, received five swipe samples on December 10, 2013 from the Northern Biomedical Research Facility in Norton Shores, Michigan. The samples were analyzed for tritium and carbon-14 according to the NRC Form 303 supplied with the samples. The sample identification numbers are presented in Table 1 and the tritium and carbon-14 results are provided in Table 2. The pertinent procedure references are included with the data tables.

  10. Fast Flux Test Facility (FFTF) standby plan

    SciTech Connect (OSTI)

    Hulvey, R.K.

    1997-03-06

    The FFTF Standby Plan, Revision 0, provides changes to the major elements and project baselines to maintain the FFTF plant in a standby condition and to continue washing sodium from irradiated reactor fuel. The Plan is consistent with the Memorandum of Decision approved by the Secretary of Energy on January 17, 1997, which directed that FFTF be maintained in a standby condition to permit the Department to make a decision on whether the facility should play a future role in the Department of Energy`s dual track tritium production strategy. This decision would be made in parallel with the intended December 1998 decision on the selection of the primary, long- term source of tritium. This also allows the Department to review the economic and technical feasibility of using the FFTF to produce isotopes for the medical community. Formal direction has been received from DOE-RL and Fluor 2020 Daniel Hanford to implement the FFTF standby decision. The objective of the Plan is maintain the condition of the FFTF systems, equipment and personnel to preserve the option for plant restart within three and one-half years of a decision to restart, while continuing deactivation work which is consistent with the standby mode.

  11. Preparatory steps for a robust dynamic model for organically bound tritium dynamics in agricultural crops

    SciTech Connect (OSTI)

    Melintescu, A.; Galeriu, D.; Diabate, S.; Strack, S.

    2015-03-15

    The processes involved in tritium transfer in crops are complex and regulated by many feedback mechanisms. A full mechanistic model is difficult to develop due to the complexity of the processes involved in tritium transfer and environmental conditions. First, a review of existing models (ORYZA2000, CROPTRIT and WOFOST) presenting their features and limits, is made. Secondly, the preparatory steps for a robust model are discussed, considering the role of dry matter and photosynthesis contribution to the OBT (Organically Bound Tritium) dynamics in crops.

  12. Field Deployable Tritium Assay System Remote Control Software

    Energy Science and Technology Software Center (OSTI)

    1998-05-12

    The FDTASREM software is a command control based application for the Field Deployable Tritium Assay System (FDTAS-Invention Disclosure SRS-96-091 has been submitted). The program runs on the Remote computer which is located at the field site with the FDTAS sampling and analysis components. The application executes commands received over the connected phone line from the operator via the FDTAS Host GUI running in the laboratory some distance away. The FDTASREM controls interface with the FDTASmore » auto sampler and the analysis systems. It tells the sampler to take a sample from a specified location and send it to the analyzer. Once the sample is sent to the analyzer, FDTASREM sequences the internal valves and pumps to deliver the sample and cocktail to the counting chamber. Once the analysis is complete, the program can execute the clean command and prepare the system for the next sample.« less

  13. Field Deployable Tritium Assay System Remote Control Software

    SciTech Connect (OSTI)

    1998-05-12

    The FDTASREM software is a command control based application for the Field Deployable Tritium Assay System (FDTAS-Invention Disclosure SRS-96-091 has been submitted). The program runs on the Remote computer which is located at the field site with the FDTAS sampling and analysis components. The application executes commands received over the connected phone line from the operator via the FDTAS Host GUI running in the laboratory some distance away. The FDTASREM controls interface with the FDTAS auto sampler and the analysis systems. It tells the sampler to take a sample from a specified location and send it to the analyzer. Once the sample is sent to the analyzer, FDTASREM sequences the internal valves and pumps to deliver the sample and cocktail to the counting chamber. Once the analysis is complete, the program can execute the clean command and prepare the system for the next sample.

  14. Tritium-target performance at RTNS-II

    SciTech Connect (OSTI)

    Heikkinen, D.W.; Logan, C.M.

    1982-01-01

    The Rotating Target Neutron Source (RTNS-II) uses a 360-keV deuteron beam and the /sup 3/He(d,n)/sup 4/He reaction to generate 14-MeV neutrons. The neutrons are used for fusion materials damage studies. The tritium target consists of a band of titanium tritide on copper alloy substrates of 23- or 50-cm diameter. During operation, the substrates are internally cooled and rotated at approx. 4000 rpm to withstand beam intensities in excess of 100 mA. Neutron production data have been accumulated for fifty-eight 23-cm and five 50-cm targets. From these data, using a non-linear least-squares fitting procedure, target performance parameters have been obtained which permit a quantitative comparison of individual targets. Average parameters are obtained for the 23- and 50-cm targets.

  15. In-line chemical sensor deployment in a tritium plant

    SciTech Connect (OSTI)

    Wright, J.S.; Hope, D.T.; Torres, R.D.; Peters, B.; Tovo, L.L.

    2015-03-15

    The Savannah River Tritium Plant (TP) relies on well understood but aging sensor technology for process gas analysis. The use of alternative sensing and detection technologies for in-line and real-time analysis would aid process control and optimization. The TP upgrading follows a 2-phase projects. In the first phase, TP sensing requirements were determined by a team of process experts. Meanwhile, Savannah River National Laboratory sensor experts identified candidate technologies and related them to the TP processing requirements. The resulting road-map links the candidate technologies to actual plant needs. In the second phase an instrument demonstration station was established within a TP glove box in order to provide accurate assessments of how a candidate sensor technology would perform in a contaminated process environment.

  16. Mobile Facility

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

    govSitesMobile Facility AMF Information Science Architecture Baseline Instruments AMF1 AMF2 AMF3 MAOS AMF Fact Sheet Images Contacts AMF Deployments McMurdo Station, Antarctica, 2015-2016 Pearl Harbor, Hawaii, to San Francisco, California, 2015 Hyytiälä, Finland, 2014 Manacapuru, Brazil, 2014 Oliktok Point, Alaska, 2013 Los Angeles, California, to Honolulu, Hawaii, 2012 Cape Cod, Massachusetts, 2012 Gan Island, Maldives, 2011 Ganges Valley, India, 2011 Steamboat Springs, Colorado, 2010

  17. Facility Type!

    Office of Legacy Management (LM)

    ITY: --&L~ ----------- srct-r~ -----------~------~------- if yee, date contacted ------------- cl Facility Type! i I 0 Theoretical Studies Cl Sample 84 Analysis ] Production 1 Diepasal/Storage 'YPE OF CONTRACT .--------------- 1 Prime J Subcontract&- 1 Purchase Order rl i '1 ! Other information (i.e., ---------~---~--~-------- :ontrait/Pirchaee Order # , I C -qXlJ- --~-------~~-------~~~~~~ I I ~~~---~~~~~~~T~~~ FONTRACTING PERIODi IWNERSHIP: ,I 1 AECIMED AECMED GOVT GOUT &NTtiAC+OR

  18. Safety overview of the National Ignition Facility

    SciTech Connect (OSTI)

    Brereton, S.J.; McLouth, L.; Odell, B.; Singh, M.; Tobin, M.; Trent, M.

    1996-05-23

    The National Ignition Facility (NIF) is a proposed US Department of Energy inertial confinement laser fusion facility. The candidate sites for locating the NIF are: Los Alamos National Laboratory, Sandia National Laboratory, the Nevada Test Site, and Lawrence Livermore National Laboratory (LLNL), the preferred site. The NIF will operate by focusing 192 laser beams onto a tiny deuterium- tritium target located at the center of a spherical target chamber. The NIF mission is to achieve inertial confinement fusion (ICF) ignition, access physical conditions in matter of interest to nuclear weapons physics, provide an above ground simulation capability for nuclear weapons effects testing, and contribute to the development of inertial fusion for electrical power production. The NIF has been classified as a radiological, low hazard facility on the basis of a preliminary hazards analysis and according to the DOE methodology for facility classification. This requires that a safety analysis be prepared under DOE Order 5481.1B, Safety Analysis and Review System. A draft Preliminary Safety Analysis Report (PSAR) has been written, and this will be finalized later in 1996. This paper summarizes the safety issues associated with the operation of the NIF. It provides an overview of the hazards, estimates maximum routine and accidental exposures for the preferred site of LLNL, and concludes that the risks from NIF operations are low.

  19. Safety overview of the National Ignition Facility

    SciTech Connect (OSTI)

    Brereton, S.; McLouth, L.; Odell, B.; Singh, M.; Tobin, M.; Trent, M.; Yatabe, J.

    1996-12-31

    The National Ignition Facility (NIF) is a proposed U.S. Department of Energy inertial confinement laser fusion facility. The candidate sites for locating the NIF are: Los Alamos National Laboratory, Sandia National Laboratory - New Mexico, the Nevada Test Site, and Lawrence Livermore National Laboratory (LLNL), the preferred site. The NIF will operate by focusing 192 laser beams onto a tiny deuterium-tritium target located at the center of a spherical target chamber. The NIF has been classified as a radiological, low hazard facility on the basis of a preliminary hazards analysis and according to the DOE methodology for facility classification. This requires that a safety analysis be prepared under DOE Order 5481.1B, Safety Analysis and Review System. A draft Preliminary Safety Analysis Report (PSAR) has been written, and this will be finalized later in 1996, after independent review. This paper summarizes the safety issues associated with the construction and operation of the NIF. It provides an overview of the hazards, estimates maximum routine and accidental exposures for the preferred site of LLNL, and concludes that the risks from NIF operations are low. 9 refs., 2 figs., 2 tabs.

  20. Savannah River Site | National Nuclear Security Administration | (NNSA)

    National Nuclear Security Administration (NNSA)

    Savannah River Site NNSA operates facilities at the Savannah River Site to supply and process tritium, a radioactive form of hydrogen that is a key component of nuclear weapons. SRS loads tritium and non-tritium reservoirs; including reclamation of previously used tritium reservoirs, processing of reservoirs; recycling, extraction, and enrichment of tritium gas and lab operations. SRS also plays a critical role in NNSA's nonproliferation missions. SRS is run by Savannah River Nuclear Solutions.

  1. ScaLAPACK

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

    Savannah River Site NNSA operates facilities at the Savannah River Site to supply and process tritium, a radioactive form of hydrogen that is a key component of nuclear weapons. SRS loads tritium and non-tritium reservoirs; including reclamation of previously used tritium reservoirs, processing of reservoirs; recycling, extraction, and enrichment of tritium gas and lab operations. SRS also plays a critical role in NNSA's nonproliferation missions. SRS is run by Savannah River Nuclear Solutions.

  2. Fast flux test facility radioisotope production and medical applications

    SciTech Connect (OSTI)

    Schenter, R.E.; Smith, S.G.; Tenforde, T.S.

    1997-12-01

    The Fast Flux Test Facility (FFTF) is a 400-MW, sodium-cooled reactor that operated successfully from 1982 to 1992, conducting work in support of the liquid-metal reactor industry by developing and testing fuel assemblies, control rods, and other core reactor components. Upon termination of this program, the primary mission of FFTF ended, and it was placed in a standby mode in 1993. However, in January 1997 the U.S. Secretary of Energy requested that FFTF be evaluated for a future mission that would consist of a primary goal of producing tritium for nuclear defense applications and a secondary goal of supplying medical isotopes for research and clinical applications. Production by FFTF of tritium for U.S. nuclear weapons would augment the dual-track strategy now under consideration for providing a long-term tritium supply in the United States (consisting of a light water reactor option and an accelerator option). A decision by the Secretary of Energy on proceeding with steps leading toward the possible reactivation of FFTF will be made before the end of 1998.

  3. Research Facilities | NREL

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

    Research Facilities Our state-of-the-art facilities are available to industry entrepreneurs, engineers, scientists, and universities for researching and developing their energy technologies. Our researchers and technicians who operate these labs and facilities are ready to work with you and share their expertise. Alphabetical Listings Laboratories Test and User Facilities Popular Facilities Energy Systems Integration Facility Integrated Biorefinery Research Facility Process Development

  4. Finding of No Significant Impact for the Storage of Tritium-Producing...

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

    for the Storage of Tritium-Producing Burnable Absorber Rods in K-Area Transfer Bay at the Savannah River Site Agency: U.S. Department of Energy (DOE) Action: Finding of No...

  5. Modeling and validating tritium transfer in a grassland ecosystem in response to {sup 3}H releases

    SciTech Connect (OSTI)

    Le Dizes, S.

    2015-03-15

    In this paper a radioecological model (TOCATTA) for tritium transfer in a grassland ecosystem developed on an hourly time-step basis is proposed and compared with the first data set obtained in the vicinity of the AREVA-NC reprocessing plant of La Hague (France). The TOCATTA model aims at simulating dynamics of tritium transfer in agricultural soil and plant ecosystems exposed to time-varying HTO concentrations in air water vapour and possibly in irrigation and rain water. In the present study, gaseous releases of tritium from the AREVA NC nuclear reprocessing plant in normal operation can be intense and intermittent over a period of less than 24 hours. A first comparison of the model predictions with the field data has shown that TOCATTA should be improved in terms of kinetics of tritium transfer.

  6. Celebrating the 20th anniversary of the tritium shot heard around...

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

    were so good and so thorough that the tritium shots were successful early on in the D-T campaign." The preparations mobilized physicists, engineers and staffers throughout the...

  7. Potential Emissions of Tritium in Air from Wells on the Nevada National Security Site

    SciTech Connect (OSTI)

    Warren, R.

    2012-10-08

    This slide-show discusses the Nevada National Security Site (NNSS) and tritium in the groundwater. It describes the wells and boreholes and potential airflow from these sources. Monitoring of selected wells is discussed and preliminary results are presented.

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

    SciTech Connect (OSTI)

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

    2008-07-15

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

  9. Target designs for Accelerator Production of Tritium (APT) utilizing lithium-aluminum

    SciTech Connect (OSTI)

    Todosow, M.; Van Tuyle, G.J.

    1996-03-01

    A number of accelerator-driven spallation neutron-source target/blanket systems have been developed for production of tritium under the APT Program. The two systems described in this paper employ a proton linear accelerator, and a target which contains a heavy-metal(s) for the production of neutrons via spallation, and solid lithium-aluminum for the production of tritium via neutron capture. lie lithium-aluminum technology is based on that employed at Savannah River for tritium production since the 1950`s. In the APT concept tritium is produced without the presence of fissionable materials; therefore, no high-level waste is produced, and the ES&H concerns are significantly reduced compared to reactor systems.

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

    Office of Environmental Management (EM)

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

  11. NREL: Research Facilities - Webmaster

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

    Your name: Your email address: Your message: Send Message Printable Version Research Facilities Home Laboratories Test & User Facilities Laboratories & Facilities by Technology...

  12. Facilities | Bioenergy | NREL

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

    Facilities At NREL's state-of-the-art bioenergy research facilities, researchers design ... facility to develop, test, evaluate, and demonstrate bioenergy processes and technologies. ...

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

    SciTech Connect (OSTI)

    Chan, V. S.; Stambaugh, R

    2011-01-01

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

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

    SciTech Connect (OSTI)

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

    2011-01-01

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

  15. Research Facility,

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

    Delivering the Data As a general condition for use of the ARM Climate Research Facility, users are required to include their data in the ARM Data Archive. All data acquired must be of sufficient quality to be useful and must be documented such that users will be able to clearly understand the meaning and organization of the data. Final, quality-assured data sets are stored in the Data Archive and are freely accessible to the general scientific community. Upon conclusion of the field campaign,

  16. Support - Facilities - Radiation Effects Facility / Cyclotron...

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

    During experiments at the Radiation Effects Facility users are assisted by the experienced ... shops are available to the users of the Radiation Effects Facility for design, ...

  17. Radiation Effects Facility - Facilities - Cyclotron Institute

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

    Radiation Effects Facility Typical DUT(device under test) set-up at the end of the Radiation Effects beamline. The Radiation Effects Facility is available for commercial, ...

  18. Harrisburg Facility Biomass Facility | Open Energy Information

    Open Energy Info (EERE)

    2006 Database Retrieved from "http:en.openei.orgwindex.php?titleHarrisburgFacilityBiomassFacility&oldid397545" Feedback Contact needs updating Image needs updating...

  19. Brookhaven Facility Biomass Facility | Open Energy Information

    Open Energy Info (EERE)

    2006 Database Retrieved from "http:en.openei.orgwindex.php?titleBrookhavenFacilityBiomassFacility&oldid397235" Feedback Contact needs updating Image needs updating...

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

    SciTech Connect (OSTI)

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

    1992-12-31

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