Sample records for tritium extraction facility

  1. Structural acceptance criteria Remote Handling Building Tritium Extraction Facility

    SciTech Connect (OSTI)

    Mertz, G.

    1999-12-16T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Lewis, M.R.

    2000-01-11T23:59:59.000Z

    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.

  3. STAR Facility Tritium Accountancy

    SciTech Connect (OSTI)

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

    2007-09-01T23:59:59.000Z

    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.

  4. STAR facility tritium accountancy

    SciTech Connect (OSTI)

    Pawelko, R. J.; Sharpe, J. P.; Denny, B. J. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415 (United States)

    2008-07-15T23:59:59.000Z

    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)

  5. Weapons engineering tritium facility overview

    SciTech Connect (OSTI)

    Najera, Larry [Los Alamos National Laboratory

    2011-01-20T23:59:59.000Z

    Materials provide an overview of the Weapons Engineering Tritium Facility (WETF) as introductory material for January 2011 visit to SRS. Purpose of the visit is to discuss Safety Basis, Conduct of Engineering, and Conduct of Operations. WETF general description and general GTS program capabilities are presented in an unclassified format.

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

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

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

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

    Energy Savers [EERE]

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

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

    SciTech Connect (OSTI)

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

    1997-11-14T23:59:59.000Z

    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.

  9. Closing the TSTA Facility, tritium removed from TSTA

    SciTech Connect (OSTI)

    Tesch, Charles; Rogers, M. L. (Michael L.); Michelotti, R. A. (Roy A.)

    2004-01-01T23:59:59.000Z

    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 determine the tritium inventory. The removed tritium inventory was either sent to other facilities for processing or buried at the Los Alamos radioactive waste disposal site. No effort was made to recover tritiated water absorbed on molecular sieve. Some hardware was sent to other facilities for reuse. One complete experimental system, including a contaminated glovebox and many components, was packaged and transferred to another DOE site for future use. Special burial containers that could safely contain up to 10 grams of tritium per package were designed and fabricated. The entire project was conducted with low tritium emission to the environment and negligible personnel exposure. After completion of the tritium removal, all remaining hardware and piping were opened and vented, and facility emission was below 1 Ci per day.

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

    SciTech Connect (OSTI)

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

    2004-09-01T23:59:59.000Z

    The purpose of this paper is to present the current status of the development of the Safety and Tritium Applied Research (STAR) Facility at the Idaho National Engineering and Environmental Laboratory (INEEL). Designated a National User Facility by the US DOE, the primary mission of STAR is to provide laboratory infrastructure to study tritium science and technology issues associated with the development of safe and environmentally friendly fusion energy. Both tritium and non-tritium fusion safety research is pursued along three key thrust areas: (1) plasma-material interactions of plasma-facing component (PFC) materials exposed to energetic tritium and deuterium ions, (2) fusion safety concerns related to PFC material chemical reactivity and dust/debris generation, activation product mobilization, and tritium behavior in fusion systems, and (3) molten salts and fusion liquids for tritium breeder and coolant applications. STAR comprises a multi-room complex with operations segregated to permit both tritium and non-tritium activities in separately ventilated rooms. Tritium inventory in STAR is limited to 15,000 Ci to maintain its classification as a Radiological Facility. Experiments with tritium are typically conducted in glovebox environments. Key components of the tritium infrastructure have been installed and tested. This includes the following subsystems: (1) a tritium Storage and Assay System (SAS) that uses two 50-g depleted uranium beds for tritium storage and PVT/beta-scintillation analyses for tritium accountability measurements, (2) a Tritium Cleanup System (TCS) that uses catalytic oxidation and molecular sieve water absorption to remove tritiated species from glovebox atmosphere gases and gaseous effluents from experiment and process systems, and (3) tritium monitoring instrumentation for room air, glovebox atmosphere and stack effluent tritium concentration measurements. Integration of the tritium infrastructure subsystems with the experimental and laboratory process systems is planned for early in 2004. Following an operational readiness review, tritium operations will be initiated in the summer of 2004. Summary results of the performance testing of the tritium infrastructure subsystems and their integration into the laboratory operations will be presented at this conference. Current research activity includes plasma-material interaction studies with the Tritium Plasma Experiment (TPE) and tritium/chemistry interactions in the molten salt designated as Flibe (2·LiF-BeF2). The implementation of these capabilities in STAR will be described.

  11. Estimate of Legacy Tritium in Building 232-H Tritium Facility, Savannah River Site

    SciTech Connect (OSTI)

    Clark, E.A.

    2003-01-07T23:59:59.000Z

    This report describes an estimate of how much tritium will be held up in those parts of the 232-H process that will remain in the building after deactivation The anticipated state of this tritium is also discussed. This information will be used to assess the radiological status of the deactivated facility.

  12. Health physics manual of good practices for tritium facilities

    SciTech Connect (OSTI)

    Blauvelt, R.K.; Deaton, M.R.; Gill, J.T. [and others

    1991-12-01T23:59:59.000Z

    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.

  13. Radiological Characterization and Final Facility Status Report Tritium Research Laboratory

    SciTech Connect (OSTI)

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

    1996-08-01T23:59:59.000Z

    This document contains the specific radiological characterization information on Building 968, the Tritium Research Laboratory (TRL) Complex and Facility. We performed the characterization as outlined in its Radiological Characterization Plan. The Radiological Characterization and Final Facility Status Report (RC&FFSR) provides historic background information on each laboratory within the TRL complex as related to its original and present radiological condition. Along with the work outlined in the Radiological Characterization Plan (RCP), we performed a Radiological Soils Characterization, Radiological and Chemical Characterization of the Waste Water Hold-up System including all drains, and a Radiological Characterization of the Building 968 roof ventilation system. These characterizations will provide the basis for the Sandia National Laboratory, California (SNL/CA) Site Termination Survey .Plan, when appropriate.

  14. Analysis of tritium extraction from liquid lithium by permeation window and solid gettering processes

    SciTech Connect (OSTI)

    Takeda, T. [Japan Atomic Energy Research Inst., Ibaraki-ken (Japan); Ying, A.Y.; Abdou, M.A. [Univ. of California, Los Angeles, CA (United States)

    1994-12-31T23:59:59.000Z

    Tritium recovery from liquid lithium at low concentration is an important problem for liquid metal breeder-blanket in a fusion reactor. Previous studies have identified tritium recovery methods including molten salt extraction, gettering recovery, permeation window, and vacuum distillation. In this paper, the authors focus on the numerical studies on tritium extraction by permeation window and gettering processes. These studies include for example: dynamic tritium concentration variation along the flow direction, tritium inventory distributions in the permeator and getter bed, along with the effect of dispersion on extraction efficiency. Using a model description makes it possible to determine functional dependence and provide insight into the interrelationships of the various operating conditions and material properties which may affect the behavior of tritium in the material. Clearly, reliable material properties (such as diffusivity, solubility, etc.) are essential for realistic evaluations.

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

    Broader source: Energy.gov [DOE]

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

  16. NNSA TRITIUM SUPPLY CHAIN

    SciTech Connect (OSTI)

    Wyrick, Steven [Savannah River National Laboratory, Aiken, SC, USA; Cordaro, Joseph [Savannah River National Laboratory, Aiken, SC, USA; Founds, Nanette [National Nuclear Security Administration, Albuquerque, NM, USA; Chambellan, Curtis [National Nuclear Security Administration, Albuquerque, NM, USA

    2013-08-21T23:59:59.000Z

    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.

  17. 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.

    1997-08-01T23:59:59.000Z

    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.

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

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels DataDepartment of Energy Your Density Isn't YourTransport(Fact Sheet),EnergyImprovement ofDecemberPlateau RemediationofFacility -

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

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels DataDepartment of Energy Your Density Isn't YourTransport(Fact Sheet),EnergyImprovement ofDecemberPlateau RemediationofFacility -2012 |

  20. Plutonium Uranium Extraction Facility Documented Safety Analysis

    SciTech Connect (OSTI)

    DODD, E.N.

    2003-10-08T23:59:59.000Z

    This document provides the documented safety analysis (DSA) and Central Plateau Remediation Project (CP) requirements that apply to surveillance and maintenance (S&M) activities at the Plutonium-Uranium Extraction (PUREX) facility. This DSA was developed in accordance with DOE-STD-1120-98, ''Integration of Environment, Safety, and Health into Facility Disposition Activities''. Upon approval and implementation of this document, the current safety basis documents will be retired.

  1. Summary of Topic1 Fusion Power Extraction

    E-Print Network [OSTI]

    Abdou, Mohamed

    Extraction and Tritium Fuel Cycle · What choices are available for material, coolant, breeder, configuration availability of external tritium supply? #12;FW/Blanket concepts for fusion power extraction and tritium&D and facilities strongly overlap RAFM Steel PbLi Breeder Helium Cooled Ceramic Breeder Beryllium Helium Cooled Pb

  2. Tritium monitor

    DOE Patents [OSTI]

    Chastagner, Philippe (Augusta, GA)

    1994-01-01T23:59:59.000Z

    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.

  3. Tritium monitor

    DOE Patents [OSTI]

    Chastagner, P.

    1994-06-14T23:59:59.000Z

    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.

  4. 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

    2009-09-01T23:59:59.000Z

    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.

  5. Natural phenomena risk analysis - an approach for the tritium facilities 5480.23 SAR natural phenomena hazards accident analysis

    SciTech Connect (OSTI)

    Cappucci, A.J. Jr.; Joshi, J.R.; Long, T.A.; Taylor, R.P.

    1997-07-01T23:59:59.000Z

    A Tritium Facilities (TF) Safety Analysis Report (SAR) has been developed which is compliant with DOE Order 5480.23. The 5480.23 SAR upgrades and integrates the safety documentation for the TF into a single SAR for all of the tritium processing buildings. As part of the TF SAR effort, natural phenomena hazards (NPH) were analyzed. A cost effective strategy was developed using a team approach to take advantage of limited resources and budgets. During development of the Hazard and Accident Analysis for the 5480.23 SAR, a strategy was required to allow maximum use of existing analysis and to develop a cost effective graded approach for any new analysis in identifying and analyzing the bounding accidents for the TF. This approach was used to effectively identify and analyze NPH for the TF. The first part of the strategy consisted of evaluating the current SAR for the RTF to determine what NPH analysis could be used in the new combined 5480.23 SAR. The second part was to develop a method for identifying and analyzing NPH events for the older facilities which took advantage of engineering judgment, was cost effective, and followed a graded approach. The second part was especially challenging because of the lack of documented existing analysis considered adequate for the 5480.23 SAR and a limited budget for SAR development and preparation. This paper addresses the strategy for the older facilities.

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

    SciTech Connect (OSTI)

    Wiegand, D.L.

    1994-09-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Greager, E.M.

    1997-12-11T23:59:59.000Z

    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.

  8. Tritium monitoring techniques

    SciTech Connect (OSTI)

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

    1996-05-01T23:59:59.000Z

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

  9. Report on the oversight assessment of the operational readiness review of the Replacement Tritium Facility at Savannah River Site

    SciTech Connect (OSTI)

    Lee, B.T.

    1993-03-01T23:59:59.000Z

    This report presents the results of an oversight assessment (OA) conducted by the US Department of Energy's (DOE) Office of Environment, Safety and Health (EH) of operational readiness review (ORR) activities for the Replacement Tritium Facility (RTF) located at Savannah River Site (SRS). The EH OA of this facility took place concurrently with an ORR conducted by the DOE Office of Defense Programs (DP). The DP ORR was conducted from January 19 through February 5, 1993. The EH OA was performed in accordance with the protocol and procedures specified in EH Program for Oversight Assessment of Operational Readiness Evaluations for Startups and Restarts,'' dated September 15, 1992. The EH OA Team evaluated the DP ORR to determine whether it was thorough and demonstrated sufficient inquisitiveness to verify that the implementation of programs and procedures adequately ensures the protection of worker safety and health. The EH OA Team performed its evaluation of the DP ORR in the following technical areas: occupational safety, industrial hygiene, and respiratory protection; fire protection; and chemical safety. In the areas of fire protection and chemical safety, the EH OA Team conducted independent vertical-slice reviews to confirm DP ORR results. Within each technical area, the EH OA Team reviewed the DP ORR Plan, including the Criteria Review and Approach Documents (CRADs); the qualifications of individual DP ORR team members; the performance of planned DP ORR activities; and the results of the DP ORR.

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

    E-Print Network [OSTI]

    2006-01-01T23:59:59.000Z

    of Tritium in Engineered and Earth Materials Progress Reportof Tritium in Engineered and Earth Materials Progress Reportof Tritium in Engineered and Earth Materials Progress Report

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

    E-Print Network [OSTI]

    2006-01-01T23:59:59.000Z

    A. , and S. Childress, Tritium production in the Dolomitic4. Figure 5.3-5. Mobility of Tritium in Engineered and EarthInverse Henry’s constant + Tritium half-life 1 . 0 × 10 ? 5

  12. 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.

    1996-10-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Steimke, J.L.

    1999-04-06T23:59:59.000Z

    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.

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

    National Nuclear Security Administration (NNSA)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn AprilA Approved: 5-13-14Russianvolunteer | National011-03-2010EIS NewsJuly 2014N I T E D

  15. Tritium assay of Li sub 2 O pellets in the LBM/LOTUS experiments

    SciTech Connect (OSTI)

    Quanci, J.; Azam, S.; Bertone, P.

    1986-01-01T23:59:59.000Z

    One of the objectives of the Lithium Blanket Module (LBM) program is to test the ability of advanced neutronics codes to model the tritium breeding characteristics of a fusion blanket exposed to a toroidal fusion neutron source. The LBM consists of over 20,000 cylindrical lithium oxide pellets and numerous diagnostic pellets and wafers. The LBM has been irradiated at the Ecole Polytechnique Federale de Lausanne (EPFL) LOTUS facility with a Haefely sealed neutron generator that gives a point deuterium-tritium neutron source up to 5 {times} 10{sup 12} 14-MeV n/s. Both Princeton Plasma Physics Laboratory (PPL) and EPFL assayed the tritium bred at various positions in the LBM. EPFL employed a dissolution technique while PPL recovered the tritium by a thermal extraction method. EPFL uses 0.38-g, 75% TD, lithium oxide diagnostic wafers to evaluate the tritium bred in the LBM. PPPL employs a thermal extraction method to determine the tritium bred in lithium oxide samples. In the initial experiments, diagnostic pellets and wafers were placed at five locations in the LBM central removable test rod at distances of 3, 9, 21, 36, and 48 cm from the front face of the module. The two sets of data for the tritium bred in the LBM along its centerline as a function of distance from the front face of the module were compared with each other, and with the predictions of two-dimensional neutronics codes. 1 ref.

  16. 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

    2014-10-01T23:59:59.000Z

    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.

  17. Tritium Systems Test Assembly (TSTA) Stabilization

    SciTech Connect (OSTI)

    Tesch, Chuck; Carlson, Richard; Michelotti, Roy; Rogers, Mike; Willms, Scott [Los Alamos National Laboratory (United States)

    2005-07-15T23:59:59.000Z

    The Los Alamos National Laboratory (LANL) 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. In 2001, the US Department of Energy (DOE) determined that TSTA's mission was complete and that the facility should be stabilized.At the completion of the stabilization project in 2003, TSTA was categorized as a radiological facility. Before stabilization was complete, the 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 of these, tritium content was characterized, a path for removal was determined, and the proper disposal package was developed. Hydrogen exchange, calorimetry, direct sampling, pressure/composition/temperature, radiological smear surveys, and controlled regeneration were used to determine the tritium inventory for each category of tritium.After removal, the tritium inventory was either (1) sent to other facilities for reuse processing or (2) buried at the LANL radioactive waste disposal site. One complete experimental system was packaged and transferred to another DOE site for future use. Special burial containers were designed and fabricated for the inventory buried at the LANL radioactive waste disposal site. The project was conducted with low tritium emission to the environment and negligible personnel exposure. After the tritium removal was complete, all remaining hardware and piping were opened and vented; the facility emission was below 1 Ci per day.

  18. Tritium systems test assembly stabilization

    SciTech Connect (OSTI)

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

    2004-01-01T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    2006-01-01T23:59:59.000Z

    shutting down the ventilation system is also shown. Thethrough the ventilation system. Tritium concentrations inby two engineered ventilation systems, and the exhaust air

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

    E-Print Network [OSTI]

    2006-01-01T23:59:59.000Z

    tritium transport in porous materials (concrete, rock) andsaturated concrete during drying, Trans. Porous Media , 24,porous medium given the diffusivity in free water. The concrete

  1. Development of New Tritium Labelling Methods for Peptides

    E-Print Network [OSTI]

    Development of New Tritium Labelling Methods for Peptides & Investigation of Guest-Host Mediated parts of the work presented here is Part I; which have involved the installation of a Tritium Chemistry Facility for the synthesis of radiolabelled compounds with tritium, and Part II; the development of new

  2. Independent Oversight Review, Savannah River Field Office Tritium...

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

    River Site (SRS) tritium facilities implemented at the activity-level by Savannah River Nuclear Solutions, LLC and its subcontractors. The review was performed by the...

  3. The control of tritium in ETHEL

    SciTech Connect (OSTI)

    Housiadas, C.; Perujo, A.; Vassallo, G. [Safety Technology Institute, Ispra (Italy)

    1994-03-01T23:59:59.000Z

    The operation of the European Tritium Handling Experimental Laboratory (ETHEL) will require the implementation of means and procedures for allowing tritium control within the facility. For that purpose, account must be taken of the particular characteristics of tritium, such as its high mobility, capacity to dissolve in materials, often limited precision when performing inventory measurements. This paper estimates the influence of these effects on the overall tritium balance in ETHEL. By employing available models for predicting tritium hold-up, it is estimated that three to four grams of tritium may potentially remain irreversibly fixed in various plant items of the standard laboratory infrastructure (exclusive of experimental circuits). On the other hand, the highest overall precision that may be attained with the present plant regarding inventory measurements is estimated to be of the order of few percent. On the basis of the above estimates, the allowable limits for the Material Unaccounted For (MUF) are discussed.

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

    SciTech Connect (OSTI)

    Lohrasbi, J.; Johnson, D.L. [Westinghouse Hanford Co., Richland, WA (United States); De Lorenzo, D.S. [Los Alamos Technical Associates, NM (United States)

    1993-12-01T23:59:59.000Z

    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.

  5. 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.

    1997-07-24T23:59:59.000Z

    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.

  6. Measurement and monitoring of tritium and other critical issues in Lead Lithium Ceramic Breeder (LLCB)

    SciTech Connect (OSTI)

    Tangri, V. K.; Mohan, S. [Heavy Water Div., Bhabha Atomic Research Centre (India); Narayanan, A.; Narayan, K. K. [Radiation Safety Systems Div., Bhabha Atomic Research Centre (India)

    2008-07-15T23:59:59.000Z

    A new Indian concept involving a lead lithium ceramic breeder is being explored. LLCB based tritium blanket modules require tritium extraction from lead-lithium as well as from helium purge gas. This paper addresses the concept of efficiency enhancement using high surface area, low-pressure drop structured gas liquid contactors for tritium extraction from the lead lithium. Conceptual flow schemes for both loops are discussed and critical issues are highlighted. Tritium monitoring systems (TMS) for measurement and monitoring of tritium is also dealt. A fast responding tritium monitor has also been developed for in situ measurement of tritium in water or gas form. It has been tested for liquid effluents. (authors)

  7. 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...

  8. Tritium Management Control of tritium inventory is

    E-Print Network [OSTI]

    Princeton Plasma Physics Laboratory

    Tritium Management · Control of tritium inventory is fundamental to public acceptance of fusion. Tritium Inventory Buildup vs. retention rate 1% 3% 10% 20% 50% TFTR JET TS ITER Inventory limit 10 days Production Rates (best estimate) carbon beryllium tungsten carbon with flakes DustInventory(kg) Number

  9. Continuous aqueous tritium monitor

    DOE Patents [OSTI]

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

    1989-05-30T23:59:59.000Z

    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.

  10. Radiological Training for Tritium Facilities

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn April 23, 2014, an OHASeptember 2010In addition to 1 |DDOE HDBK-1113-2008 April

  11. Radiological Training for Tritium Facilities

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn April 23, 2014, an OHASeptember 2010In addition to 1 |DDOE HDBK-1113-2008 AprilDOE

  12. REMOTE ANALYSIS OF HIGH-TRITIUM-CONTENT WATER

    SciTech Connect (OSTI)

    Diprete, D; Raymond Sigg, R; Leah Arrigo, L; Donald Pak, D

    2007-08-07T23:59:59.000Z

    Systems to safely analyze for tritium in moisture collected from glovebox atmospheres are being developed for use at Savannah River Site (SRS) tritium facilities. Analysis results will guide whether the material contains sufficient tritium for economical recovery, or whether it should be stabilized for disposal as waste. In order to minimize potential radiation exposures that could occur in handling and diluting high-tritium-content water, SRS sought alternatives to the process laboratory's routine analysis by liquid-scintillation counting. The newer systems determine tritium concentrations by measuring bremsstrahlung radiation induced by low-energy beta interactions. One of the systems determines tritium activity in liquid streams, the other determines tritium activity in water vapor. Topics discussed include counting results obtained by modeling and laboratory testing and corrections that are made for low-energy photon attenuation.

  13. Continuous aqueous tritium monitor

    DOE Patents [OSTI]

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

    1987-10-19T23:59:59.000Z

    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.

  14. TRITIUM SYSTEMS KEYWORDS: tritium fuel cycle, re-

    E-Print Network [OSTI]

    Abdou, Mohamed

    FUSION REACTORS WILLIAM KUAN and MOHAMED A. ABDOU* University of California at Los Angeles, School in detail by Abdou et al.1 The value Lr is the required tritium breeding ratio ~TBR!, and La a simplified first-order linear system model, which made use of mean tritium residence times*E-mail: abdou

  15. PDRD (SR13046) TRITIUM PRODUCTION FINAL REPORT

    SciTech Connect (OSTI)

    Smith, P.; Sheetz, S.

    2013-09-30T23:59:59.000Z

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

  16. REPORT OF SURVEY OF THE LOS ALAMOS TRITIUM SYSTEMS TEST ASSEMBLY...

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

    connection, exterior to the facility. The main experimental area contains tritium and depleted uranium (the latter in hydride storage beds). There are 15 glove boxes in which...

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

    E-Print Network [OSTI]

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

  18. Accounting strategy of tritium inventory in the heavy water detritiation pilot plant from ICIT Rm. Valcea

    SciTech Connect (OSTI)

    Bidica, N.; Stefanescu, I. [Inst. of Cryogenics and Isotopes Technologies, Uzinei Str. No. 4, Rm. Valcea (Romania); Cristescu, I. [TLK, Forschungszentrum Karlsruhe, Postfach 3640, D76021 Karlsruhe (Germany); Bornea, A.; Zamfirache, M.; Lazar, A.; Vasut, F.; Pearsica, C.; Stefan, I. [Inst. of Cryogenics and Isotopes Technologies, Uzinei Str. No. 4, Rm. Valcea (Romania); Prisecaru, I.; Sindilar, G. [Univ. Politehnica of Bucharest, Splaiul Independentei 313, Bucharest (Romania)

    2008-07-15T23:59:59.000Z

    In this paper we present a methodology for determination of tritium inventory in a tritium removal facility. The method proposed is based on the developing of computing models for accountancy of the mobile tritium inventory in the separation processes, of the stored tritium and of the trapped tritium inventory in the structure of the process system components. The configuration of the detritiation process is a combination of isotope catalytic exchange between water and hydrogen (LPCE) and the cryogenic distillation of hydrogen isotopes (CD). The computing model for tritium inventory in the LPCE process and the CD process will be developed basing on mass transfer coefficients in catalytic isotope exchange reactions and in dual-phase system (liquid-vapour) of hydrogen isotopes distillation process. Accounting of tritium inventory stored in metallic hydride will be based on in-bed calorimetry. Estimation of the trapped tritium inventory can be made by subtraction of the mobile and stored tritium inventories from the global tritium inventory of the plant area. Determinations of the global tritium inventory of the plant area will be made on a regular basis by measuring any tritium quantity entering or leaving the plant area. This methodology is intended to be applied to the Heavy Water Detritiation Pilot Plant from ICIT Rm. Valcea (Romania) and to the Cernavoda Tritium Removal Facility (which will be built in the next 5-7 years). (authors)

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

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011 Strategic2Uranium Transferon theTedRegion | Department of Energy4th TritiumFacility |

  20. Tritium Plasma Experiment and

    Office of Environmental Management (EM)

    September 23-25, 2014 3 Figures Implantation Recombination Diffusion Traps or lattice damage Surface damage and erosion Trapping Detrapping Permeation tritium atom lattice atom...

  1. Tritium Technology at CNL

    Office of Environmental Management (EM)

    -11- UNRESTRICTED ILLIMIT * Use of tritium decay-energy as power source * Convert high-energy electrons into low-voltage current * Two methods: * Indirect: beta photons ...

  2. 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)

    NONE

    1996-04-01T23:59:59.000Z

    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.

  3. Irradiation Testing of Blanket Materials at the HFR Petten with On Line Tritium Monitoring

    SciTech Connect (OSTI)

    Magielsen, A.J.; Laan, J.G. van der; Hegeman, J.B.J.; Stijkel, M.P.; Ooijevaar, M.A.G

    2005-07-15T23:59:59.000Z

    Irradiation experiments are performed in support of fusion blanket technology development. These comprise ceramic solid breeder materials, and a liquid Lithium Lead alloy, as well as blanket subassemblies and components. Experimental facilities at the HFR to study tritium release, permeation characteristics, and neutron irradiation performance, have recently been extended. This paper gives an overview on the tritium breeding materials irradiation programme and describes the facilities required for irradiation testing and on-line tritium measurement.

  4. Methods for tritium labeling

    DOE Patents [OSTI]

    Andres, Hendrik (Hochwald, CH); Morimoto, Hiromi (El Cerrito, CA); Williams, Philip G. (Oakland, CA)

    1993-01-01T23:59:59.000Z

    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.

  5. 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.

    1998-11-10T23:59:59.000Z

    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.

  6. Oxidative Tritium Decontamination System

    DOE Patents [OSTI]

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

    2006-02-07T23:59:59.000Z

    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.

  7. FINAL REPORT FOR TRITIUM WATER MONITOR

    SciTech Connect (OSTI)

    Sigg, R.; Ferguson, B.; DiPrete, D.

    2011-04-25T23:59:59.000Z

    The objective of this Plant Directed Research and Demonstration (PDRD) task was to develop a system to safetly analyze tritium in moisture collected from glovebox atmospheres in the Savannah River Site (SRS) Tritium Facility. In order to minimize potential radiation exposures that could occur in handling and diluting high-tritium-content water, SRS sought alternatives to liquid-scintillation counting. The proposed system determines tritium concentrations by measuring Bremsstrahlung radiation induced by low-energy beta interactions in liquid samples. Results show that, after a short counting period (30 seconds), detection limits are three orders of magnitude below the described concentration of tritiated water in the zeolite beds. Additionally, this report covers the analysis of process samples and the investigation of several cell window materials including beryllium, aluminum, and copper. Final tests reveal that alternate window materials and thicknesses can be used to obtain useful results. In particular, a window of stainless steel of moderate thickness (0.3 cm) can be used for counting relatively high levels of tritium.

  8. Effect of Sawtooth Activity on Tritium and Beam Deuterium Evolution in Trace Tritium Experiments on JET

    E-Print Network [OSTI]

    Effect of Sawtooth Activity on Tritium and Beam Deuterium Evolution in Trace Tritium Experiments on JET

  9. Five years of tritium handling experience at the Tritium Systems Test Assembly

    SciTech Connect (OSTI)

    Carlson, R.V.

    1989-01-01T23:59:59.000Z

    The Tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory is a facility designed to develop and demonstrate, in full scale, technologies necessary for safe and efficient operation of tritium systems required for tokamak fusion reactors. TSTA currently consists of systems for evacuating reactor exhaust gas with compound cryopumps; for removing impurities from plasma exhaust gas and recovering the chemically-combined tritium; for separating the isotopes of hydrogen; for transfer pumping; or storage of hydrogen isotopes; for gas analysis; and for assuring safety by the necessary control, monitoring, and tritium removal from effluent streams. TSTA also has several small scale experiments to develop and test new equipment and processes necessary for fusion reactors. In this paper, data on component reliability, failure types and rates, and waste quantities are presented. TSTA has developed a Quality Assurance program for preparing and controlling the documentation of the procedures required for the design, purchase, and operation of the tritium systems. Operational experience under normal, abnormal, and emergency conditions is presented. One unique aspect of operations at TSTA is that the design personnel for the TSTA systems are also part of the operating personnel. This has allowed for the relatively smooth transition from design to operations. TSTA has been operated initially as a research facility. As the system is better defined, operations are proceeding toward production modes. The DOE requirements for the operation of a tritium facility like TSTA include personnel training, emergency preparedness, radiation protection, safety analysis, and preoperational appraisals. The integration of these requirements into TSTA operations is discussed. 4 refs., 3 figs., 3 tabs.

  10. Tritium waste package

    DOE Patents [OSTI]

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

    1995-11-07T23:59:59.000Z

    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.

  11. Tritium breeding blanket

    SciTech Connect (OSTI)

    Smith, D.; Billone, M.; Gohar, Y. (Argonne National Lab., IL (USA)); Baker, C. (Oak Ridge National Lab., TN (USA)); Mori, S.; Kuroda, T.; Maki, K.; Takatsu, H.; Yoshida, H. (Japan Atomic Energy Research Inst., Tokyo (Japan)); Raffray, A. (California Univ., Los Angeles, CA (USA)); Sviatoslavsky, I. (Wisconsin Univ., Madison, WI (USA)); Simbolotti, G. (ENEA, Frascati (Italy). Centro Ricerche Energia); Dae

    1991-01-01T23:59:59.000Z

    The terms of reference for ITER provide for incorporation of a tritium breeding blanket with a breeding ratio as close to unity as practical. A breeding blanket is required to assure an adequate supply of tritium to meet the program objectives. Based on specified design criteria, a ceramic breeder concept with water coolant and an austenitic steel structure has been selected as the first option and lithium-lead blanket concept has been chosen as an alternate option. The first wall, blanket, and shield are integrated into a single unit with separate cooling systems. The design makes extensive use of beryllium to enhance the tritium breeding ratio. The design goals with a tritium breeding ratio of 0.8--0.9 have been achieved and the R D requirements to qualify the design have been identified. 4 refs., 8 figs., 2 tabs.

  12. Tritium waste package

    DOE Patents [OSTI]

    Rossmassler, Rich (Cranbury, NJ); Ciebiera, Lloyd (Titusville, NJ); Tulipano, Francis J. (Teaneck, NJ); Vinson, Sylvester (Ewing, NJ); Walters, R. Thomas (Lawrenceville, NJ)

    1995-01-01T23:59:59.000Z

    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.

  13. 2009 EVALUATION OF TRITIUM REMOVAL AND MITIGATION TECHNOLOGIES FOR WASTEWATER TREATMENT

    SciTech Connect (OSTI)

    LUECK KJ; GENESSE DJ; STEGEN GE

    2009-02-26T23:59:59.000Z

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

  14. MODELING ATMOSPHERIC RELEASES OF TRITIUM FROM NUCLEAR INSTALLATIONS

    SciTech Connect (OSTI)

    Okula, K

    2007-01-17T23:59:59.000Z

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

  15. Drum bubbler tritium processing system

    DOE Patents [OSTI]

    Rule, Keith (Hopewell, NJ); Gettelfinger, Geoff (Lexington, MA); Kivler, Paul (Hamilton Square, NJ)

    1999-01-01T23:59:59.000Z

    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.

  16. Tritium deposition patterns in TFTR

    E-Print Network [OSTI]

    Princeton Plasma Physics Laboratory

    Tritium deposition patterns in TFTR Presented by C. H. Skinner with key contributions from Charles, JAERI #12;· TFTR was a limiter machine - no divertor. · Operated with tritium Nov `93 - April `97. · NetV Limiter Temperature @ 28 MW NBI Low density, high temperature edge #12;Tritium deposition patterns in TFTR

  17. Monitoring of tritium

    DOE Patents [OSTI]

    Corbett, James A. (Turtle Creek, PA); Meacham, Sterling A. (Greensburg, PA)

    1981-01-01T23:59:59.000Z

    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.

  18. Drum bubbler tritium processing system

    DOE Patents [OSTI]

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

    1999-08-17T23:59:59.000Z

    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.

  19. 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.

    2012-01-09T23:59:59.000Z

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

  20. Tritium APEX Interim Report November, 1999

    E-Print Network [OSTI]

    California at Los Angeles, University of

    Tritium APEX Interim Report November, 1999 16-1 CHAPTER 16: TRITIUM Contributors Lead Author: Dai Kai Sze #12;Tritium APEX Interim Report November, 1999 16-2 16. TRITIUM 16.1 Design constraints Tritium recovery and containment are some of the key issues associated with breeding blanket design

  1. UK contractors' experience of management of tritium during decommissioning projects

    SciTech Connect (OSTI)

    Green, Tommy; Stevens, Keith; Heaney, John [NUKEM Ltd., Kelburn Court, Daten Park, Birchwood, Warrington, WA3 6TW (United Kingdom); Murray, Alan [Tetronics Limited, 1 Ram Court. Wicklesham Farm, Faringdon. Oxfordshire SN7 7PN (United Kingdom); Warwick, Phil; Croudace, Ian [GAU-Radioalytical, National Oceanography Centre (United Kingdom)

    2007-07-01T23:59:59.000Z

    Available in abstract form only. Full text of publication follows: This paper provides an account of the tritium management experience of a UK decommissioning and remediation contracting organisation (NUKEM Limited), supported by a specialist radio-analysis organisation (GAU-Radioanalytical). This experience was gained during the execution of projects which involved the characterisation and remediation of facilities which had previously been used for tritium work and were contaminated with tritium. The emphasis of the paper is on the characterisation (sampling and analysis) of tritium. An account is given of the development of a methodology to improve the accuracy of tritium characterisation. The improved methodology evolved from recognition of the need to minimise tritium losses during sampling, storage, transport and preparation for analysis. These improvements were achieved in a variety of ways, including use of cold and dry sampling techniques in preference to hot or wet ones and freezing relevant samples during storage and transport. The major benefit was an improvement in the accuracy and reliability of the analyses results, essential for proper categorisation, sentencing and future management of tritiated waste. (authors)

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

    SciTech Connect (OSTI)

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

    1983-01-01T23:59:59.000Z

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

  3. Process for the removal of tritium from the product solutions obtained by the Purex process

    SciTech Connect (OSTI)

    Bossche, A.V.; Olinger, R.

    1983-02-22T23:59:59.000Z

    A process for the removal of tritium from the product solutions obtained in the reprocessing of irradiated nuclear fuels by the Purex process comprising a plurality of series-connected extraction cycles having an organic solvent.

  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)

    1992-01-01T23:59:59.000Z

    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. 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.

    1992-01-14T23:59:59.000Z

    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.

  6. Welding tritium aged stainless steel

    SciTech Connect (OSTI)

    Kanne, W.R. Jr.

    1993-04-01T23:59:59.000Z

    Stainless steels exposed to tritium become unweldable by conventional methods due to He buildup within the metal matrix. With longer service lives expected for new weapon systems, and service life extensions of older systems, methods for welding/repair on tritium-exposed material will become important. Results are reported that indicate that both solid-state resistance welding and low-heat gas metal arc overlay welding are promising methods for repair or modification of tritium-aged stainless steel.

  7. Producing tritium in a homogenous reactor

    DOE Patents [OSTI]

    Cawley, William E. (Richland, WA)

    1985-01-01T23:59:59.000Z

    A method and apparatus are described for the joint production and separation of tritium. Tritium is produced in an aqueous homogenous reactor and heat from the nuclear reaction is used to distill tritium from the lower isotopes of hydrogen.

  8. Tritium monitoring of groundwater and surfaces

    SciTech Connect (OSTI)

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

    1999-03-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    S. A. Bruyere; L. C. Cadwallader

    2011-09-01T23:59:59.000Z

    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.

  10. 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.

    2014-03-06T23:59:59.000Z

    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.

  11. Chapter 4. Uranium Mine and Extraction Facility Reclamation This chapter is not intended to serve as guidance, or to supplement EPA or other agency environmental

    E-Print Network [OSTI]

    4-1 Chapter 4. Uranium Mine and Extraction Facility Reclamation This chapter is not intended, it is an outline of practices which may or have been used for uranium site restoration. Mining reclamation for uranium mining sites. The existence of bonding requirements and/or financial guarantees in the cases where

  12. Tritium-field betacells

    SciTech Connect (OSTI)

    Walko, R.J.; Lincoln, R.C.; Baca, W.E. (Sandia National Labs., Albuquerque, NM (USA)); Goods, S.H. (Sandia National Labs., Livermore, CA (USA)); Negley, G.H. (AstroPower, Inc., Newark, DE (USA))

    1991-01-01T23:59:59.000Z

    Betavoltaic power sources operate by converting the nuclear decay energy of beta-emitting radioisotopes into electricity. Since they are not chemically driven, they could operate at temperatures which would either be to hot or too cold for typical chemical batteries. Further, for long lived isotopes, they offer the possibility of multi-decade active lifetimes. Two approaches are being investigated: direct and indirect conversion. Direct conversion cells consist of semiconductor diodes similar to photovoltaic cells. Beta particle directly bombard these cells, generating electron-hole pairs in the semiconductor which are converted to useful power. Many using low power flux beta emitters, wide bandgap semiconductors are required to achieve useful conversion efficiencies. The combination of tritium, as the beta emitter, and gallium phosphide (GaP), as the semiconductor converter, was evaluated. Indirect conversion betacells first convert the beta energy to light with a phosphor, and then to electricity with photovoltaic cells. An indirect conversion power source using a tritium radioluminescent (RL) light is being investigated. Our analysis indicates that this approach has the potential for significant volume and cost savings over the direct conversion method. 7 refs., 11 figs.

  13. Tritium at Fermilab Fermilab Community Advisory Board

    E-Print Network [OSTI]

    Quigg, Chris

    Tritium at Fermilab Fermilab Community Advisory Board September 23, 2010 Rob Plunkett, Fermilab #12 and other equipment. #12;4 Fermilab, before Nov. 2005 Tritium is produced as part of accelerator operations: Tritium detected Our routine testing of surface water at Fermilab revealed low levels of tritium: 3

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

    SciTech Connect (OSTI)

    Mutch Jr, R. D. [HydroQual, Inc., 1200 MacArthur Blvd., Mahwah, NJ 07430 (United States); Manhattan College, Riverdale, NY (United States); Columbia Univ., New York, NY (United States); Mahony, J. D. [HydroQual, Inc., 1200 MacArthur Blvd., Mahwah, NJ 07430 (United States); Manhattan College, Riverdale, NY (United States)

    2008-07-15T23:59:59.000Z

    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)

  15. 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.

    1988-02-01T23:59:59.000Z

    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.

  16. Composition containing aerogel substrate loaded with tritium

    DOE Patents [OSTI]

    Ashley, Carol S. (Albuquerque, NM); Brinker, C. Jeffrey (Albuquerque, NM); Ellefson, Robert E. (Centerville, OH); Gill, John T. (Miamisburg, OH); Reed, Scott (Albuquerque, NM); Walko, Robert J. (Albuquerque, NM)

    1992-01-01T23:59:59.000Z

    The invention provides a process for loading an aerogel substrate with tritium and the resultant compositions. According to the process, an aerogel substrate is hydrolyzed so that surface OH groups are formed. The hydrolyzed aerogel is then subjected to tritium exchange employing, for example, a tritium-containing gas, whereby tritium atoms replace H atoms of surface OH groups. OH and/or CH groups of residual alcohol present in the aerogel may also undergo tritium exchange.

  17. Differential atmospheric tritium sampler

    DOE Patents [OSTI]

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

    1987-10-02T23:59:59.000Z

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

  18. Differential atmospheric tritium sampler

    DOE Patents [OSTI]

    Griesbach, Otto A. (Langhorne, PA); Stencel, Joseph R. (Skillman, NJ)

    1990-01-01T23:59:59.000Z

    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.

  19. Tritium Determination at Trace Level: Which Strategy to Determine Accurately HTO and OBT in Environmental Samples?

    SciTech Connect (OSTI)

    Baglan, N.; Alanic, G.; Pointurier, F. [CEA (France)

    2005-07-15T23:59:59.000Z

    Focusing on environmental tritium levels, measurements have been made for several natural water and leaf samples from an area where no tritium industrial discharge was known to occur. Therefore, to obtain sufficiently accurate data tritium was determined from large samples. Moreover, tritium measurement at environmental level requires appropriate methodology to avoid any contamination. Both tissue free water tritium (TFWT) and organically bound tritium (OBT) were determined for biological samples and compared to preliminary or literature data. In this paper, the authors describe both: a mobile extraction water device allowing therefore to realise the extraction step under the sampling site conditions to get rid of any contamination; a sensitive method for low level non-exchangeable OBT determination by a combination of a suitable sample treatment, a large capacity combustion apparatus and low background liquid scintillation spectrometry. Then, owing to validate this approach the authors give an application of this methodology to the determination of both fractions determined on tree leaves originating from the lower Rhone valley. The results demonstrate both the suitability of the procedure as tritium concentration in leaves and natural waters exhibit environmental level concentration and also that the OBT background in the studied area is very close to the one measured in the south west of France.

  20. Primer on tritium safe handling practices

    SciTech Connect (OSTI)

    Not Available

    1994-12-01T23:59:59.000Z

    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.

  1. Laser-assisted isotope separation of tritium

    DOE Patents [OSTI]

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

    1983-01-01T23:59:59.000Z

    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.

  2. Revised per Referee's comments October 20, 1999 HEAT DEPOSITION, DAMAGE, AND TRITIUM BREEDING

    E-Print Network [OSTI]

    California at Los Angeles, University of

    that can efficiently extract power from fusion devices with high neutron wall load. Among the concepts@fusion.ucla.edu Abstract The Advanced Power Extraction (APEX) study aims at exploring new and innovative blanket concepts of the liquid FW/B/S and protection of the VV and magnet against radiation damage, (2) profiles of tritium

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

    E-Print Network [OSTI]

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

  4. 8 2. Helium und Tritium in der Geosphre 2. Helium und Tritium in der Geosphre

    E-Print Network [OSTI]

    Aeschbach-Hertig, Werner

    8 2. Helium und Tritium in der Geosphäre 2. Helium und Tritium in der Geosphäre 2.1. Spezielle Einheiten und Konstanten An dieser Stelle sollen die speziellen für Helium und Tritium verwendeten Einheiten definiert und dazugehörige Umrechnungen angegeben werden. Die Wahl der Werte einiger für Helium und Tritium

  5. Overview of Recent Tritium Experiments in TPE

    SciTech Connect (OSTI)

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

    2010-10-01T23:59:59.000Z

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

  6. Neutrino mass limit from tritium beta decay

    E-Print Network [OSTI]

    E. W. Otten; C. Weinheimer

    2009-09-11T23:59:59.000Z

    The paper reviews recent experiments on tritium beta spectroscopy searching for the absolute value of the electron neutrino mass $m(\

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

    SciTech Connect (OSTI)

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

    2011-09-26T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2012-02-29T23:59:59.000Z

    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.

  9. Assessment of molecular effects on neutrino mass measurements from tritium beta decay

    E-Print Network [OSTI]

    Bodine, L I; Robertson, R G H

    2015-01-01T23:59:59.000Z

    The beta decay of molecular tritium currently provides the highest sensitivity in laboratory-based neutrino mass measurements. The upcoming Karlsruhe Tritium Neutrino (KATRIN) experiment will improve the sensitivity to 0.2 eV, making a percent-level quantitative understanding of molecular effects essential. The modern theoretical calculations available for neutrino-mass experiments agree with spectroscopic data. Moreover, when neutrino-mass experiments performed in the 1980s with gaseous tritium are re-evaluated using these modern calculations, the extracted neutrino mass-squared values are consistent with zero instead of being significantly negative. On the other hand, the calculated molecular final-state branching ratios are in tension with dissociation experiments performed in the 1950s. We re-examine the theory of the final-state spectrum of molecular tritium decay and its effect on the determination of the neutrino mass, with an emphasis on the role of the vibrational- and rotational-state distribution i...

  10. Tritium Recovery from Solid Breeder Blanket by Water Vapor Addition to Helium Sweep Gas

    SciTech Connect (OSTI)

    Kawamura, Yoshinori; Iwai, Yasunori; Nakamura, Hirofumi; Hayashi, Takumi; Yamanishi, Toshihiko; Nishi, Masataka [Japan Atomic Energy Research Institute (Japan)

    2005-07-15T23:59:59.000Z

    In the solid breeder blanket of fusion reactor, bred tritium is planned to be extracted from the blanket as HT by passing of H{sub 2}-added sweep gas in general. In that case, tritium leakage by permeation to coolant can not be ignored. So, the application of H{sub 2}O-added sweep gas is discussed, with which tritium leakage to coolant can be much reduced. As the result of discussion, H{sub 2}O-added sweep gas is probable method of tritium recovery. For the further detailed discussion, it is important to enrich the data correlated to the interaction of H{sub 2}, H{sub 2}O, breeder, multiplier and structures.

  11. EFFECTS OF TRITIUM GAS EXPOSURE ON POLYMERS

    SciTech Connect (OSTI)

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

    2011-01-07T23:59:59.000Z

    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.

  12. Tritium hazard via the ingestion pathway

    SciTech Connect (OSTI)

    Travis, C.C.

    1985-01-01T23:59:59.000Z

    The classic methodology for estimating dose to man from environmental tritium ignores the fact that organically bound tritium in foodstuffs may be directly assimilated in the bound compartment of tissues without previous oxidation. We propose a four-compartment model that allows for the ability to input organically bound tritium in foodstuffs directly into the organic compartments of the model. We found that organically bound tritium in foodstuffs can increase the total body dose by a factor of 1.7 to 4.5 times the free body water dose alone, depending on the bound to loose ratio of tritium in the diet. 10 refs., 1 fig., 1 tab.

  13. Tritium Supply Considerations Scott Willms

    E-Print Network [OSTI]

    to refurbish/restart K New Production Reactor project start -MHTGH, HWR, LWR · 1990 Ebasco HWR and MHTGR of tritium · Old DOE price was $10,000/gm · Present Canada price is ~$30,000/gm · Expected cost for future US

  14. Tritium pellet injector TPI-1

    SciTech Connect (OSTI)

    Viniar, I.V.; Kuteev, B.V.; Koblents, P.Yu. [Technical Univ., Saint-Petersburg (Russian Federation); Saksagansky, G.L.; Skripunov, V.N. [Efremov Inst., Saint-Petersburg (Russian Federation)

    1995-12-31T23:59:59.000Z

    The current concept of fueling large fusion devices is based on gas puffing and pellet injection. The pellet injector produces, accelerates and transports into a plasmas the pellets composed of hydrogen isotopes. Here, tasks and design of a tritium repeating pellet injector developed in Russia are presented. The injector uses improved in-situ and extrusion technologies for pellet formation.

  15. Tritium recovery from carbon particulate Until 2009 the JET machine has operated with a

    E-Print Network [OSTI]

    objectives Design and construction of a facility to recover tritium from carbon. Including: · Commissioning of the material. case study DT fuel cycle Solution Significant R&D effort went into developing an oxidation: technologyservices@ccfe.ac.uk www.ccfe.ac.uk/technologyservices.aspx Recovery system during construction #12;

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

    SciTech Connect (OSTI)

    Pruess, Karsten; Conrad, Mark; Finsterle, Stefan; Kennedy, Mack; Kneafsey, Timothy; Salve, Rohit; Su, Grace; Zhou, Quanlin

    2006-10-25T23:59:59.000Z

    This report details the work done between June 13 andSeptember 30, 2006 by Lawrence Berkeley National Laboratory (LBNL)scientists to assist Fermi National Accelerator Laboratory (Fermilab)staff in understanding tritium transport at the Neutrino at the MainInjector (NuMI) facility. As a byproduct of beamline operation, thefacility produces (among other components) tritium in engineeredmaterials and the surrounding rock formation. Once the tritium isgenerated, it may be contained at the source location, migrate to otherregions within the facility, or be released to theenvironment.

  17. DEPLOYMENT OF THE BULK TRITIUM SHIPPING PACKAGE

    SciTech Connect (OSTI)

    Blanton, P.

    2013-10-10T23:59:59.000Z

    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.

  18. Tritium radioluminescent devices, Health and Safety Manual

    SciTech Connect (OSTI)

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

    1995-06-01T23:59:59.000Z

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

  19. Recovery of tritium from tritiated molecules

    DOE Patents [OSTI]

    Swansiger, W.A.

    1984-10-17T23:59:59.000Z

    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.

  20. TRITIUM INCORPORATION STUDIES IN PHOTO-SYNTHETIC BACTERIA

    E-Print Network [OSTI]

    Dehner, Thomas R.; Chan, W.-S.; Caple, Marianne B.; Calvin, M.

    2008-01-01T23:59:59.000Z

    W-7405-eng-48 UCRL-17749 I TRITIUM INCORPORATION STUDIES INand M. Calvin July 1967 TRITIUM INCORPOWION STUDIES I Ntransport, w are studying t h e tritium labeling pattern e i

  1. Experimental studies of the transfer phenomena of tritium in an isotope exchange column for recovery tritium

    E-Print Network [OSTI]

    Experimental studies of the transfer phenomena of tritium in an isotope exchange column for recovery tritium Anisia Bornea, Ion Cristescu, Marius Zamfirache, Carmen Varlam National Institute of R processes for tritium separation, is the catalyst isotope exchange water-hydrogen. The main problem

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

    National Nuclear Security Administration (NNSA)

    Production and Environmental Impacts The production of tritium in a commercial light water reactor (CLWR) is technically straightforward. Most existing CLWRs utilize 12-foot-long...

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

    National Nuclear Security Administration (NNSA)

    Tritium Enterprise exceeds productivity savings goals for FY13 | National Nuclear Security Administration Facebook Twitter Youtube Flickr RSS People Mission Managing the Stockpile...

  4. Thermal Removal Of Tritium From Concrete And Soil To Reduce Groundwater Impacts

    SciTech Connect (OSTI)

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

    2012-12-04T23:59:59.000Z

    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 of four units to batch treat concrete rubble and soil. Post treatment sampling verified that the activity in the treated soil and concrete met the treatment standards for each medium which allowed the treated concrete rubble and soil to be disposed of on site as backfill. During testing and operations a total of 1,261-m{sup 3} (1,650-yd{sup 3}) of contaminated concrete and soils were treated with an actual incurred cost of $3,980,000. This represents a unit treatment cost of $3,156/m{sup 3} ($2,412/yd{sup 3}). In 2011 the project was recognized with an e-Star Sustainability Award by DOE's Office of Environmental Management.

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

    SciTech Connect (OSTI)

    Jackson, Dennis G. [Savannah River National Laboratory, Building 773-42A, Aiken, South Carolina 29808 (United States)] [Savannah River National Laboratory, Building 773-42A, Aiken, South Carolina 29808 (United States); Blount, Gerald C. [Savannah River Nuclear Solutions (United States)] [Savannah River Nuclear Solutions (United States); Wells, Leslie H.; Cardoso, Joao E.; Kmetz, Thomas F.; Reed, Misty L. [U.S Department of Energy-Savannah River Site (United States)] [U.S Department of Energy-Savannah River Site (United States)

    2013-07-01T23:59:59.000Z

    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 of four units to batch treat concrete rubble and soil. Post treatment sampling verified that the activity in the treated soil and concrete met the treatment standards for each medium which allowed the treated concrete rubble and soil to be disposed of on-site as backfill. During testing and operations a total of 1,261-m{sup 3} (1,650-yd{sup 3}) of contaminated concrete and soils were treated with an actual incurred cost of $3,980,000. This represents a unit treatment cost of $3,156/m{sup 3} ($2,412/yd{sup 3}). In 2011 the project was recognized with an e-Star Sustainability Award by DOE's Office of Environmental Management. (authors)

  6. Tritium research laboratory cleanup and transition project final report

    SciTech Connect (OSTI)

    Johnson, A.J.

    1997-02-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    2013-07-01T23:59:59.000Z

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

  8. TRITIUM EFFECTS ON WELDMENT FRACTURE TOUGHNESS

    SciTech Connect (OSTI)

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

    2006-07-17T23:59:59.000Z

    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''.

  9. 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

    2001-08-24T23:59:59.000Z

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

  10. Vanadium hydride deuterium-tritium generator

    DOE Patents [OSTI]

    Christensen, Leslie D. (Livermore, CA)

    1982-01-01T23:59:59.000Z

    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.

  11. DOE handbook: Tritium handling and safe storage

    SciTech Connect (OSTI)

    NONE

    1999-03-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    2003-02-27T23:59:59.000Z

    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. The derived concentration guidelines, combined with the tritium removal system that was developed for this project, provided a significant timesaving for decontamination as well as an overall cost savings for waste disposal.

  13. Investigation of the tritium release from Building 324 in which the stack tritium sampler was off, April 14 through 17, 1998

    SciTech Connect (OSTI)

    Brown, D.H.

    1998-06-30T23:59:59.000Z

    On April 14, 1998, a Pacific Northwest National Laboratory (PNNL) researcher performing work in the Building 324 facility approached facility management and asked if facility management could turn off the tritium sampler in the main exhaust stack. The researcher was demonstrating the feasibility of treating components from dismantled nuclear weapons in a device called a plasma arc furnace and was concerned that the sampler would compromise classified information. B and W Hanford Company (BWHC) operated the facility, and PNNL conducted research as a tenant in the facility. The treatment of 200 components in the furnace would result in the release of up to about 20 curies of tritium through the facility stack. The exact quantity of tritium was calculated from the manufacturing data for the weapons components and was known to be less than 20 curies. The Notice of Construction (NOC) approved by the Washington State Department of Health (WDOH) had been modified to allow releasing 20 curies of tritium through the stack in support of this research. However, there were irregularities in the way the NOC modification was processed. The researcher was concerned that data performed on the sampler could be used to back-calculate the tritium content of the components, revealing classified information about the design of nuclear weapons. He had discussed this with the PNNZ security organization, and they had told him that data from the sampler would be classified. He was also concerned that if he could not proceed with operation of the plasma arc furnace, the furnace would be damaged. The researcher told BWHC management that the last time the furnace was shut down and restarted it had cost $0.5 million and caused a six month delay in the project`s schedule. He had already begun heating up the furnace before recognizing the security problem and was concerned that stopping the heatup could damage the furnace. The NOC that allowed the research did not have an explicit requirement to operate the sampler during a release. The sampler was installed several years previously for other research. After reviewing the NOC and other safety basis documents, and after consulting environmental compliance specialists, facility management agreed to turn off the sampler.

  14. TRITIUM RESERVOIR STRUCTURAL PERFORMANCE PREDICTION

    SciTech Connect (OSTI)

    Lam, P.S.; Morgan, M.J

    2005-11-10T23:59:59.000Z

    The burst test is used to assess the material performance of tritium reservoirs in the surveillance program in which reservoirs have been in service for extended periods of time. A materials system model and finite element procedure were developed under a Savannah River Site Plant-Directed Research and Development (PDRD) program to predict the structural response under a full range of loading and aged material conditions of the reservoir. The results show that the predicted burst pressure and volume ductility are in good agreement with the actual burst test results for the unexposed units. The material tensile properties used in the calculations were obtained from a curved tensile specimen harvested from a companion reservoir by Electric Discharge Machining (EDM). In the absence of exposed and aged material tensile data, literature data were used for demonstrating the methodology in terms of the helium-3 concentration in the metal and the depth of penetration in the reservoir sidewall. It can be shown that the volume ductility decreases significantly with the presence of tritium and its decay product, helium-3, in the metal, as was observed in the laboratory-controlled burst tests. The model and analytical procedure provides a predictive tool for reservoir structural integrity under aging conditions. It is recommended that benchmark tests and analysis for aged materials be performed. The methodology can be augmented to predict performance for reservoir with flaws.

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

    SciTech Connect (OSTI)

    Klein, J.; Estochen, E.

    2014-03-06T23:59:59.000Z

    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. Update: tritium at Fermilab Fermilab Community Advisory Board

    E-Print Network [OSTI]

    Quigg, Chris

    Update: tritium at Fermilab Fermilab Community Advisory Board January 26, 2012 Kurt Riesselmann, Fermilab #12;2 How is tritium produced? · In nature, tritium is produced when cosmic particles hit the particles in Earth's atmosphere · Tritium is also produced in small quantities in accelerator operations

  17. Version 11/5/99 Summary of Chapter 16: Tritium

    E-Print Network [OSTI]

    California at Los Angeles, University of

    Version 11/5/99 Summary of Chapter 16: Tritium The tritium recovery systems for different breeding recommended. The recovery method has to limit the tritium inventory below the deisgn limit by safety consideration. ITER set the maximum allowable releasable tritium inventory in each component to less than 200 g

  18. Tritium monitor with improved gamma-ray discrimination

    DOE Patents [OSTI]

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

    1985-01-01T23:59:59.000Z

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

  19. Tritium monitor with improved gamma-ray discrimination

    DOE Patents [OSTI]

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

    1982-10-21T23:59:59.000Z

    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.

  20. A new solid state tritium surface monitor

    SciTech Connect (OSTI)

    Willms, R. S. (Richard Scott); Dogruel, D. (David); Myers, R. (Richard); Farrell, R. (Richard)

    2004-01-01T23:59:59.000Z

    Traditionally the amount of tritium on a surface is determined by swiping the surface with a material such as filter paper and counting the removed tritium by scintillation. While effective, this method can be time consuming, can alter the surface and only measures removable tritium. For a given application each of these considerations may or may not be a disadvantage. A solid state monitor, on the other hand, has the potential to provide rapid analysis, not alter the surface and measure all tritium on a surface. This allure has promoted open wall ion chamber and PIN diode-based tritium surface monitor development, and these techniques have enjoyed certain success. Recently the first tests were performed with an avalanche photodiode (APD) for surface tritium measurement. The tritium surface concentration is determined by placing the APD within a few millimeters of the surface of interest. Beta decay from the surface tritium impacts the APD resulting in amplified current through the diode. Analysis of this signal with a multi-channel analyzer enables counting of beta decay events and determination of the beta energy spectrum. While quite similar in concept to PIN diode based measurements, side-by-side testing showed that the APD provided substantially better counting efficiency. Considerations included count rate, background, sensitivity, stability and effect of ambient light. An important factor in the U.S. for a tritium surface monitor is the ability to measure concentrations down to the 'free release' limit, i.e., the concentration below which items can be removed from radiological control areas. The two limits being used are 10,000 disintegrations per min (dpm)/100 cm{sup 2} and 1,000 dpm/100 cm{sup 2}. Present tests show that the APD is capable of measuring down to 1,000 dpm/100 cm{sup 2} in reasonable count times. Data from this promising technique will be presented in this paper.

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

    DOE Patents [OSTI]

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

    1992-03-31T23:59:59.000Z

    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.

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

    DOE Patents [OSTI]

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

    1992-01-01T23:59:59.000Z

    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.

  3. Development of Tritium Permeation Analysis Code (TPAC)

    SciTech Connect (OSTI)

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

    2010-10-01T23:59:59.000Z

    Idaho National Laboratory developed the Tritium Permeation Analysis Code (TPAC) for tritium permeation in the Very High Temperature Gas Cooled Reactor (VHTR). All the component models in the VHTR were developed and were embedded into the MATHLAB SIMULINK package with a Graphic User Interface. The governing equations of the nuclear ternary reaction and thermal neutron capture reactions from impurities in helium and graphite core, reflector, and control rods were implemented. The TPAC code was verified using analytical solutions for the tritium birth rate from the ternary fission, the birth rate from 3He, and the birth rate from 10B. This paper also provides comparisons of the TPAC with the existing other codes. A VHTR reference design was selected for tritium permeation study from the reference design to the nuclear-assisted hydrogen production plant and some sensitivity study results are presented based on the HTGR outlet temperature of 750 degrees C.

  4. EIS-0161: Tritium Supply and Recycling

    Broader source: Energy.gov [DOE]

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

  5. Tritium proof-of-principle injector experiment

    SciTech Connect (OSTI)

    Fisher, P.W.; Milora, S.L.; Combs, S.K.; Carlson, R.V.; Coffin, D.O.

    1988-01-01T23:59:59.000Z

    The Tritium Proof-of-Principle (TPOP) pellet injector was designed and built by Oak Ridge National Laboratory (ORNL) to evaluate the production and acceleration of tritium pellets for fueling future fision reactors. The injector uses the pipe-gun concept to form pellets directly in a short liquid-helium-cooled section of the barrel. Pellets are accelerated by using high-pressure hydrogen supplied from a fast solenoid valve. A versatile, tritium-compatible gas-handling system provides all of the functions needed to operate the gun, including feed gas pressure control and flow control, plus helium separation and preparation of mixtures. These systems are contained in a glovebox for secondary containment of tritium Systems Test Assembly (TSTA) at Los Alamos National Laboratory (LANL). 18 refs., 3 figs.

  6. Vanadium hydride deuterium-tritium generator

    DOE Patents [OSTI]

    Christensen, L.D.

    1980-03-13T23:59:59.000Z

    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. Computational modeling and analysis of airflow in a tritium storage room

    SciTech Connect (OSTI)

    Chen, Z. (Zukun); Konecni, S. (Snezana); Whicker, J. J. (Jeffrey J.)

    2003-01-01T23:59:59.000Z

    In this study, a commercial computational fluid dynamics (CFD) code, CFX-5.5, was utilized to assess flow field characteristics, and to simulate tritium gas releases and subsequent transport in a storage room in the tritium handling facility at Los Alamos. This study was done with mesh refinement and results compared. The results show a complex, ventilation-induced flow field with vortices, velocity gradients, and stagnant air pockets. This paper also explains the timedependent gas dispersion results. The numerical analysis method used in this study provides important information that is possible to be validated with an experimental technique of aerosol tracer measurement method frequently used at Los Alamos. Application of CFD can have a favorable impact on the design of ventilation systems and worker safety with consideration to facility costs.

  8. Determination of a tritium bioassay technique for nuclear facilities

    E-Print Network [OSTI]

    Sensintaffar, Edwin Lee

    1971-01-01T23:59:59.000Z

    . The minimum activity of these samples was 0. 00375 uCi, or about one-tenth the urine equivalent of a maximum permissible body burden. Instrumentation Solid samples were counted with the following types of radiation detectors 1. Thin window proportional...

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

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOnItemResearch >Internship Program The NIF andPointsThrustTrinityTrinity:

  10. Method and apparatus for controlling accidental releases of tritium

    DOE Patents [OSTI]

    Galloway, Terry R. [Berkeley, CA

    1980-04-01T23:59:59.000Z

    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.

    1980-04-01T23:59:59.000Z

    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. A compact tritium AMS system

    SciTech Connect (OSTI)

    Chiarappa, M L; Dingley, K H; Hamm, R W; Love, A H; Roberts, M L

    1999-09-23T23:59:59.000Z

    Tritium ({sup 3}H) is a radioisotope that is extensively utilized in biological and environmental research. For biological research, {sup 3}H is generally quantified by liquid scintillation counting requiring gram-sized samples and counting times of several hours. For environmental research, {sup 3}H is usually quantified by {sup 3}He in-growth which requires gram-sized samples and in-growth times of several months. In contrast, provisional studies at LLNL's Center for Accelerator Mass Spectrometry have demonstrated that Accelerator Mass Spectrometry (AMS) can be used to quantify {sup 3}H in milligram-sized biological samples with a 100 to 1000-fold improvement in detection limits when compared to scintillation counting. This increased sensitivity is expected to have great impact in the biological and environmental research community. However in order to make the {sup 3}H AMS technique more broadly accessible, smaller, simpler, and less expensive AMS instrumentation must be developed. To meet this need, a compact, relatively low cost prototype {sup 3}H AMS system has been designed and built based on a LLNL ion source/sample changer and an AccSys Technology, Inc. Radio Frequency Quadrupole (RFQ) linac. With the prototype system, {sup 3}/{sup 1}H ratios ranging from 1 x 10{sup -10} to 1 x 10{sup -13} have to be measured from milligram sized samples. With improvements in system operation and sample preparation methodology, the sensitivity limit of the system is expected to increase to approximately 1 x 10{sup -15}.

  13. EFFECTS OF TRITIUM GAS EXPOSURE ON ELECTRICALLY CONDUCTING POLYMERS

    SciTech Connect (OSTI)

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

    2009-12-16T23:59:59.000Z

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

  14. Former Reactor Facilities Surveillance and Maintenance and

    E-Print Network [OSTI]

    Ohta, Shigemi

    Cold and Dark (2012) Radiological soil contamination remains below building/cap (Cesium-137, Strontium-level radioactive contamination (Cesium-137, Strontium-90, tritium) #12;Former Reactor Facilities Surveillance, and 6 in April Characterizing deep VOC contamination in Industrial Park off-site Install total

  15. Analysis of tritium transport in irradiated beryllium

    SciTech Connect (OSTI)

    Cho, S.; Abdou, M.A. [Univ. of California, Los Angeles, CA (United States)

    1994-12-31T23:59:59.000Z

    Analysis of the beryllium tritium release results with simple analytical models indicated that tritium behavior in Be is not dominated by one simple mechanism, but by a combination of several mechanisms including surface processes and helium bubbles. A model was developed and the initial version of the model included tritium diffusion in the beryllium and the beryllium oxide, second order desorption at the solid/gas interface and diffusion through interconnected porosity. Fundamental data, tritium diffusion and desorption coefficients for Be and BeO, were derived from experimental data using the model. Beryllium is a metal to which one can generally apply the concepts of diffusion, solubility, surface processes and traps. Tritium transport in the irradiated beryllium is affected by processes occurring in the bulk, He bubbles, the bulk/surface and surface/gas interfaces. There are two types of solid/gas surfaces in the irradiated Be. One is the surface at the pure Be/He bubble interface where no oxide layer exists and the other is the surface at the BeO layer/purge gas interface. Although the material characteristics of the Be and BeO layer are different and have different activation barriers, the surface processes can be applied to both interfaces.

  16. Low technology high tritium breeding blanket concept

    SciTech Connect (OSTI)

    Gohar, Y.; Baker, C.C.; Smith, D.L.; Billone, M.C.; Cha, Y.S.; Clemmer, R.; Finn, P.A.; Hassanein, A.M.; Johnson, C.E.; Liu, Y.

    1987-10-01T23:59:59.000Z

    The main function of this low technology blanket is to produce the necessary tritium for INTOR operation with minimum first wall coverage. The INTOR first wall, blanket, and shield are constrained by the dimensions of the reference design and the protection criteria required for different reactor components and dose equivalent after shutdown in the reactor hall. It is assumed that the blanket operation at commercial power reactor conditions and the proper temperature for power generation can be sacrificed to achieve the highest possible tritium breeding ratio with minimum additional research and developments and minimal impact on reactor design and operation. A set of blanket evaluation criteria has been used to compare possible blanket concepts. Six areas: performance, operating requirements, impact on reactor design and operation, safety and environmental impact, technology assessment, and cost have been defined for the evaluation process. A water-cooled blanket was developed to operate with a low temperature and pressure. The developed blanket contains a 24 cm of beryllium and 6 cm of solid breeder both with a 0.8 density factor. This blanket provides a local tritium breeding ratio of approx.2.0. The water coolant is isolated from the breeder material by several zones which eliminates the tritium buildup in the water by permeation and reduces the changes for water-breeder interaction. This improves the safety and environmental aspects of the blanket and eliminates the costly process of the tritium recovery from the water. 12 refs., 13 tabs.

  17. accidental tritium assessment: Topics by E-print Network

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

    due to acute tritium releases CERN Preprints Summary: Tritium, which is used as a fuel of a D-T burning fusion reactor, is the most important radionuclide for the safety...

  18. assess tritium levels: Topics by E-print Network

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

    due to acute tritium releases CERN Preprints Summary: Tritium, which is used as a fuel of a D-T burning fusion reactor, is the most important radionuclide for the safety...

  19. Fusion Nuclear Science and Technology (FNST) Strategic Issues, challenges, and Facilities

    E-Print Network [OSTI]

    Abdou, Mohamed

    Fusion Nuclear Science and Technology (FNST) Strategic Issues, challenges, and Facilities Nuclear Science & Technology (FNST) The nuclear environment also affects Tritium Fuel Cycle separation PFC & Blanket T processing design dependent optics 3 #12;Fusion Nuclear Science and Technology

  20. Assessment of molecular effects on neutrino mass measurements from tritium beta decay

    E-Print Network [OSTI]

    L. I. Bodine; D. S. Parno; R. G. H. Robertson

    2015-02-12T23:59:59.000Z

    The beta decay of molecular tritium currently provides the highest sensitivity in laboratory-based neutrino mass measurements. The upcoming Karlsruhe Tritium Neutrino (KATRIN) experiment will improve the sensitivity to 0.2 eV, making a percent-level quantitative understanding of molecular effects essential. The modern theoretical calculations available for neutrino-mass experiments agree with spectroscopic data. Moreover, when neutrino-mass experiments performed in the 1980s with gaseous tritium are re-evaluated using these modern calculations, the extracted neutrino mass-squared values are consistent with zero instead of being significantly negative. On the other hand, the calculated molecular final-state branching ratios are in tension with dissociation experiments performed in the 1950s. We re-examine the theory of the final-state spectrum of molecular tritium decay and its effect on the determination of the neutrino mass, with an emphasis on the role of the vibrational- and rotational-state distribution in the ground electronic state. General features can be reproduced quantitatively from considerations of kinematics and zero-point motion. We summarize the status of validation efforts and suggest means for resolving the apparent discrepancy in dissociation rates.

  1. Separation phenomenon in the Windowless Gaseous Tritium Source of KATRIN

    E-Print Network [OSTI]

    Sharipov, Felix

    Separation phenomenon in the Windowless Gaseous Tritium Source of KATRIN experiment. Ternary separa- tion. In the KATRIN experiment, in order to analyze the spectrum of electrons emmited by Tritium decay, it is very important to know the concentration distribution of Tritium along the source

  2. Diss. ETH Nr. 10714 Helium und Tritium als Tracer fr

    E-Print Network [OSTI]

    Aeschbach-Hertig, Werner

    Diss. ETH Nr. 10714 Helium und Tritium als Tracer für physikalische Prozesse in Seen ABHANDLUNG zur Zürich 1994 #12;Kurzfassung ix Kurzfassung Der radioaktive Zerfall von 3H (Tritium) zu 3He mit einer Fluide aus dem Erdinnern. Helium und Tritium werden massenspektrometrisch analysiert. Im Rahmen dieser Ar

  3. 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.

  4. Tritium management in fusion synfuel designs

    SciTech Connect (OSTI)

    Galloway, T.R.

    1980-04-25T23:59:59.000Z

    Two blanket types are being studied: a lithium-sodium pool boiler and a lithium-oxide- or lithium-sodium pool boiler and a lithium-oxide- or aluminate-microsphere moving bed. For each, a wide variety of current technology was considered in handling the tritium. Here, we show the pool boiler with the sulfur-iodine thermochemical cycle first developed and now being piloted by the General Atomic Company. The tritium (T/sub 2/) will be generated in the lithium-sodium mixture where the concentration is approx. 10 ppM and held constant by a scavenging system consisting mainly of permeators. An intermediate sodium loop carries the blanket heat to the thermochemical cycle, and the T/sub 2/ in this loop is held to 1 ppM by a similar scavenging system. With this design, we have maintained blanket inventory at 1 kg of tritium, kept thermochemical cycle losses to 5 Ci/d and environmental loss to 10 Ci/d, and held total plant risk inventory at 7 kg tritium.

  5. Small system for tritium accelerator mass spectrometry

    DOE Patents [OSTI]

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

    1993-02-23T23:59:59.000Z

    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.

  6. Small system for tritium accelerator mass spectrometry

    DOE Patents [OSTI]

    Roberts, Mark L. (Livermore, CA); Davis, Jay C. (Livermore, CA)

    1993-01-01T23:59:59.000Z

    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.

  7. Tokamak fusion reactors with less than full tritium breeding

    SciTech Connect (OSTI)

    Evans, K. Jr.; Gilligan, J.G.; Jung, J.

    1983-05-01T23:59:59.000Z

    A study of commercial, tokamak fusion reactors with tritium concentrations and tritium breeding ratios ranging from full deuterium-tritium operation to operation with no tritium breeding is presented. The design basis for these reactors is similar to those of STARFIRE and WILDCAT. Optimum operating temperatures, sizes, toroidal field strengths, and blanket/shield configurations are determined for a sequence of reactor designs spanning the range of tritium breeding, each having the same values of beta, thermal power, and first-wall heat load. Additional reactor parameters, tritium inventories and throughputs, and detailed costs are calculated for each reactor design. The disadvantages, advantages, implications, and ramifications of tritium-depleted operation are presented and discussed.

  8. DEVELOPMENT OF THE BULK TRITIUM SHIPPING PACKAGING

    SciTech Connect (OSTI)

    Blanton, P.; Eberl, K.

    2008-09-14T23:59:59.000Z

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

  9. IN-SITU TRITIUM BETA DETECTOR

    SciTech Connect (OSTI)

    J.W. Berthold; L.A. Jeffers

    1998-04-15T23:59:59.000Z

    The objectives of this three-phase project were to design, develop, and demonstrate a monitoring system capable of detecting and quantifying tritium in situ in ground and surface waters, and in water from effluent lines prior to discharge into public waterways. The tritium detection system design is based on measurement of the low energy beta radiation from the radioactive decay of tritium using a special form of scintillating optical fiber directly in contact with the water to be measured. The system consists of the immersible sensor module containing the optical fiber, and an electronics package, connected by an umbilical cable. The system can be permanently installed for routine water monitoring in wells or process or effluent lines, or can be moved from one location to another for survey use. The electronics will read out tritium activity directly in units of pico Curies per liter, with straightforward calibration. In Phase 1 of the project, we characterized the sensitivity of fluor-doped plastic optical fiber to tritium beta radiation. In addition, we characterized the performance of photomultiplier tubes needed for the system. In parallel with this work, we defined the functional requirements, target specifications, and system configuration for an in situ tritium beta detector that would use the fluor-doped fibers as primary sensors of tritium concentration in water. The major conclusions from the characterization work are: A polystyrene optical fiber with fluor dopant concentration of 2% gave best performance. This fiber had the highest dopant concentration of any fibers tested. Stability may be a problem. The fibers exposed to a 22-day soak in 120 F water experienced a 10x reduction in sensitivity. It is not known whether this was due to the build up of a deposit (a potentially reversible effect) or an irreversible process such as leaching of the scintillating dye. Based on the results achieved, it is premature to initiate Phase 2 and commit to a prototype design for construction and test. Significant improvements must be made in fluor-doped fiber performance in order to use the method for in situ monitoring to verify compliance with current EPA drinking water standards. Additional Phase 1 fiber development work should be performed to increase the fluor dopant concentration above 2% until the self-absorption limit is observed. Continued fiber optimization work is expected to improve the sensitivity limits, and will enable application of the detector to verify compliance with the US EPA drinking water standard of 20,000 pico Curies per liter. However, if the need for monitoring higher levels of tritium in water at concentrations greater than 200,000 pico Curies per liter is justified, then prototype development and testing could proceed either as a Phase 2 stand-alone effort or in parallel with continued Phase 1 development work.

  10. Apparatus to recover tritium from tritiated molecules

    DOE Patents [OSTI]

    Swansiger, William A. (Livermore, CA)

    1988-01-01T23:59:59.000Z

    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.

  11. Progress in tritium retention and release modeling for ceramic breeders

    SciTech Connect (OSTI)

    Raffray, A.R.; Federici, G. [Max-Planck-Institut fuer Plasmaphysik, Muenchen (Germany)] [and others

    1994-12-31T23:59:59.000Z

    An important aspect of the design and analysis of ceramic breeder blankets is the ability to predict the phenomenological behavior of tritium in the ceramic breeder under operating reactor conditions. By understanding the behavior of tritium in such materials, analysis and accurate predictions can be made regarding the blanket tritium release and inventory which are key design issues based on safety and fuel self-sufficiency considerations. This paper highlights the progress in tritium modeling over the last decade. Key tritium transport mechanisms are briefly described along with the more recent and sophisticated models developed to help understand them. Recent experimental data are highlighted and model calibration and validation discussed. Finally, example applications to blanket cases are shown as illustration of current predictions for ceramic breeder blanket tritium inventory.

  12. Recovery of tritium from tritiated molecules

    DOE Patents [OSTI]

    Swansiger, William A. (Livermore, CA)

    1987-01-01T23:59:59.000Z

    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.

  13. Tritium inventory control in ITER Charles Skinner with key contributions from

    E-Print Network [OSTI]

    Princeton Plasma Physics Laboratory

    Tritium inventory control in ITER Charles Skinner with key contributions from Charles Gentile permitted" Tritium inventory control Worrisome issue: Once at the tritium limit there won't be any more

  14. Hydrogeology and tritium transport in Chicken Creek Canyon, Lawrence Berkeley National Laboratory, Berkeley, California

    E-Print Network [OSTI]

    Jordan, Preston D.; Javandel, Iraj

    2007-01-01T23:59:59.000Z

    2-1. Location of the tritium plume based upon 3rd quarter,locations shown. Figure 3-5. Tritium activities (pCi/L) inCanyon. "ND" indicates no tritium detected. Figure 3-6.

  15. JUPITER-II Molten Salt Flibe Research: An Update On Tritium, Mobilization and Redox Chemistry Experiments

    SciTech Connect (OSTI)

    D.A. Petti; D. A. Petti; G. R. Smolik; Michael F. Simpson; John P. Sharpe; R. A. Anderl; S. Fukada; Y. Hatano; Masanori Hara; Y. Oya; T. Terai; D.-K. Sze; S. Tanaka

    2005-05-01T23:59:59.000Z

    The second Japan/US Program on Irradiation Tests for Fusion Research (JUPITER-II) began on April 1, 2001. Part of the collaborative research centers on studies of the molten salt 2LiF2–BeF2 (also known as Flibe) for fusion applications. Flibe has been proposed as a self-cooled breeder in both magnetic and inertial fusion power plant designs over the last 25 years. The key feasibility issues associated with the use of Flibe are the corrosion of structural material by the molten salt, tritium behavior and control in the molten salt blanket system, and safe handling practices and releases from Flibe during an accidental spill. These issues are all being addressed under the JUPITER-II program at the Idaho National Laboratory in the Safety and Tritium Applied Research (STAR) facility. In this paper, we review the program to date in the area of tritium/deuterium behavior, Flibe mobilization under accident conditions and testing of Be as a redox agent to control corrosion. Future activities planned through the end of the collaboration are also presented.

  16. TRITIUM BARRIER MATERIALS AND SEPARATION SYSTEMS FOR THE NGNP

    SciTech Connect (OSTI)

    Sherman, S; Thad Adams, T

    2008-07-17T23:59:59.000Z

    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.

  17. 2012 ACCOMPLISHMENTS - TRITIUM AGING STUDIES ON STAINLESS STEELS

    SciTech Connect (OSTI)

    Morgan, M.

    2013-01-31T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Piyush Sabharwall; Carl Stoots; Hans A. Schmutz

    2013-03-01T23:59:59.000Z

    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

  19. Tritium Formation and Mitigation in High-Temperature Reactors

    SciTech Connect (OSTI)

    Piyush Sabharwall; Carl Stoots

    2012-10-01T23:59:59.000Z

    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

  20. EFFECTS OF TRITIUM GAS EXPOSURE ON EPDM ELASTOMER

    SciTech Connect (OSTI)

    Clark, E.

    2009-12-11T23:59:59.000Z

    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.

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

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011 Strategic2Uranium Transferon theTedRegion | Department of Energy4th TritiumFacility

  2. PISCES FY11 Research Highlight Tritium accumulation within the ITER vessel is expected to be dominated

    E-Print Network [OSTI]

    PISCES FY11 Research Highlight Tritium accumulation within the ITER vessel is expected vessel. Another possible technique to mitigate tritium accumulation in these codeposited surfaces

  3. E-Print Network 3.0 - activity tritium labelled Sample Search...

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

    12, 2009 Mohamed Abdou Summary: & Shield Components 4. Tritium Processing Systems 5. Remote Maintenance Components 6. Heat Transport... (feasibility) 2. Tritium Fuel...

  4. Improving cryogenic deuterium–tritium implosion performance on OMEGA

    SciTech Connect (OSTI)

    Sangster, T. C.; Goncharov, V. N.; Betti, R.; Radha, P. B.; Boehly, T. R.; Collins, T. J. B.; Craxton, R. S.; Delettrez, J. A.; Edgell, D. H.; Epstein, R.; Forrest, C. J.; Froula, D. H.; Glebov, Y. Yu.; Harding, D. R.; Hohenberger, M.; Hu, S. X.; Igumenshchev, I. V.; Janezic, R.; Kelly, J. H.; Kessler, T. J. [Laboratory for Laser Energetics, University of Rochester, 250 East River Road, Rochester, New York 14623 (United States)] [Laboratory for Laser Energetics, University of Rochester, 250 East River Road, Rochester, New York 14623 (United States); and others

    2013-05-15T23:59:59.000Z

    A flexible direct-drive target platform is used to implode cryogenic deuterium–tritium (DT) capsules on the OMEGA laser [Boehly et al., Opt. Commun. 133, 495 (1997)]. The goal of these experiments is to demonstrate ignition hydrodynamically equivalent performance where the laser drive intensity, the implosion velocity, the fuel adiabat, and the in-flight aspect ratio (IFAR) are the same as those for a 1.5-MJ target [Goncharov et al., Phys. Rev. Lett. 104, 165001 (2010)] designed to ignite on the National Ignition Facility [Hogan et al., Nucl. Fusion 41, 567 (2001)]. The results from a series of 29 cryogenic DT implosions are presented. The implosions were designed to span a broad region of design space to study target performance as a function of shell stability (adiabat) and implosion velocity. Ablation-front perturbation growth appears to limit target performance at high implosion velocities. Target outer-surface defects associated with contaminant gases in the DT fuel are identified as the dominant perturbation source at the ablation surface; performance degradation is confirmed by 2D hydrodynamic simulations that include these defects. A trend in the value of the Lawson criterion [Betti et al., Phys. Plasmas 17, 058102 (2010)] for each of the implosions in adiabat–IFAR space suggests the existence of a stability boundary that leads to ablator mixing into the hot spot for the most ignition-equivalent designs.

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

    SciTech Connect (OSTI)

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

    1991-01-01T23:59:59.000Z

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

  6. Application of Tritium Remote Control and Monitoring System (TRECAMS) to TFTR`s tritium inventory management program

    SciTech Connect (OSTI)

    Schobert, G.; Bashore, D.; Dong, J.; Diesso, M.; Mika, R. [Princeton Plasma Physics Lab., NJ (United States)

    1995-12-31T23:59:59.000Z

    TFTR has a stringent program to manage and account for its tritium inventory. In support of this a tritium inventory accounting capability has been implemented on TRECAMS. This was an ideal approach because TRECAMS is a high reliability system that monitors the necessary parameters, i.e., temperatures, pressures, valve positions, etc., to track the movement of tritium. It also has a powerful set of utilities which support such an application. This paper describes the application of TRECAMS to monitor the transfer of tritium between the Uranium Beds (UBEDs), the Tritium Gas Delivery Manifold (TGDM), 14 Tritium Use Point holding volumes, and the TFTR torus. Real time data is presented to the TFTR operators using graphical displays and trends. An event driven program automatically collects the data before and after tritium transfers, calculates differences and sums, tabulates the data and provides printed reports. The reports include summaries of tritium deliveries, bleedback operations, injections, a daily summary of delivery/bleedback activities, and a daily summary of injection activities. All reference data is archived and can be reproduced in a plotted or tabular format. This data can be displayed or printed by the TFTR Shift Supervisor`s VAX workstation or by anyone with an account on the laboratory`s VAX cluster.

  7. Overview of Transport, Fast Particle and Heating and Current Drive Physics using Tritium in JET plasmas

    E-Print Network [OSTI]

    Overview of Transport, Fast Particle and Heating and Current Drive Physics using Tritium in JET plasmas

  8. New Safety and Technical Challenges and Operational Experience on the JET First Trace Tritium Experiment

    E-Print Network [OSTI]

    New Safety and Technical Challenges and Operational Experience on the JET First Trace Tritium Experiment

  9. Overview of Transport, Fast Particle and Heating and Current Drive Physics using Tritium in JET Plasmas

    E-Print Network [OSTI]

    Overview of Transport, Fast Particle and Heating and Current Drive Physics using Tritium in JET Plasmas

  10. Fusion Engineering and Design 81 (2006) 11311144 Physics and technology conditions for attaining tritium

    E-Print Network [OSTI]

    Abdou, Mohamed

    2006-01-01T23:59:59.000Z

    tritium self-sufficiency for the DT fuel cycle M.E. Sawana,, M.A. Abdoub a Fusion Technology Institute online 27 December 2005 Abstract There is no practical external source of tritium for fusion energy development beyond ITER and all subsequent fusion systems have to breed their own tritium. To ensure tritium

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

    E-Print Network [OSTI]

    American Society for Testing and Materials. Philadelphia

    2009-01-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2010-10-14T23:59:59.000Z

    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.

  13. Summary of benchmark experiments for simulation of fusion reactors using an annular blanket with a line deuterium-tritium source

    SciTech Connect (OSTI)

    Maekawa, H.; Abdou, M.A.; Oyama, Y. [Japan Atomic Energy Research Inst., Ibaraki (Japan)] [and others

    1995-09-01T23:59:59.000Z

    The Japan Atomic Energy Research lnstitute (JAERI)/U.S. Department of Energy collaborative program was performed using the Fusion Neutronics Source facility at JAERI. In Phase III of this program, tritium breeding measurements were conducted in prototypical blankets driven by a simulated deuterium-tritium neutron line source. This phase differed from the earlier two phases in respect to the spatial distribution of the source as the earlier experiments were done with a point neutron source. This series basically consisted of an annular test blanket and a pseudoline source to investigate the effect of source spread on the neutronic performance. A concise description is on the outlines of the simulated line source, the test blanket systems for Phases-IIIA, -IIIB, and -IIIC, measured items, experimental results, and their analyses. 23 refs., 8 figs., 3 tabs.

  14. LANSCE | Facilities

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

    LINAC Outreach Affiliations Visiting LANSCE Facilities Isotope Production Facility Lujan Neutron Scattering Center MaRIE Proton Radiography Ultracold Neutrons Weapons Neutron...

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

    DOE Patents [OSTI]

    Anderson, D.F.

    1981-01-27T23:59:59.000Z

    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.

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

    DOE Patents [OSTI]

    Anderson, David F. (Los Alamos, NM)

    1984-01-01T23:59:59.000Z

    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.

  17. Tritium Formation and Mitigation in High Temperature Reactors

    SciTech Connect (OSTI)

    Piyush Sabharwall; Carl Stoots

    2012-08-01T23:59:59.000Z

    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.

  18. Japan Atomic Energy Research Institute/United States Integral Neutronics Experiments and Analyses for tritium breeding, nuclear heating, and induced radioactivity

    SciTech Connect (OSTI)

    Abdou, M.A.; Youssef, M.; Kumar, A. [Univ. of California, Los Angeles, CA (United States)] [and others

    1995-08-01T23:59:59.000Z

    A large member of integral experiments for fusion blanket neutronics were performed using deuterium-tritium (D-T) neutrons at the Fusion Neutronics Source facility as part of a 10-yr collaborative program between the Japan Atomic Energy Research Institute and the United States. A number of measurement techniques were developed for tritium production, induced radioactivity, and nuclear heating. Transport calculations were performed using three-dimensional Monte Carlo and two-dimensional discrete ordinates codes and the latest nuclear data libraries in Japan and the United States. Significant differences among measurement techniques and calculation methods were found. To assure a 90% confidence level for tritium breeding calculations not to exceed measurements, designers should use a safety factor > 1.1 to 1.2, depending on the calculation method. Such a safety factor may not be affordable with most candidate blanket designs. Therefore, demonstration of tritium self-sufficiency is recommended as a high priority for testing in near-term fusion facilities such as the International Thermonuclear Experimental Reactor (ITER). The radioactivity measurements were performed for > 20 materials with the focus on gamma emitters with half-lives < 5 yr. Most discrepancies were attributed directly to deficiencies in the activation libraries, particularly errors in cross sections for certain reactions. 71 refs., 30 figs., 5 tabs.

  19. 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.

    2013-10-30T23:59:59.000Z

    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.

  20. Ignition of deuterium-tritium fuel targets

    DOE Patents [OSTI]

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

    1991-08-27T23:59:59.000Z

    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.

  1. On-line tritium production monitor

    DOE Patents [OSTI]

    Mihalczo, J.T.

    1993-11-23T23:59:59.000Z

    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.

  2. On-line tritium production monitor

    DOE Patents [OSTI]

    Mihalczo, John T. (Oak Ridge, TN)

    1993-01-01T23:59:59.000Z

    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.

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

    E-Print Network [OSTI]

    Monreal, Benjamin

    2009-01-01T23:59:59.000Z

    The shape of the beta decay energy distribution is sensitive to the mass of the electron neutrino. Attempts to measure the endpoint shape of tritium decay have so far seen no distortion from the zero-mass form, thus placing an upper limit of m_nu_beta < 2.3 eV. Here we show that a new type of electron energy spectroscopy could improve future measurements of this spectrum and therefore of the neutrino mass. We propose to detect the coherent cyclotron radiation emitted by an energetic electron in a magnetic field. For mildly relativistic electrons, like those in tritium decay, the relativistic shift of the cyclotron frequency allows us to extract the electron energy from the emitted radiation. We present calculations for the energy resolution, noise limits, high-rate measurement capability, and systematic errors expected in such an experiment.

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

    E-Print Network [OSTI]

    Benjamin Monreal; Joseph A. Formaggio

    2009-04-18T23:59:59.000Z

    The shape of the beta decay energy distribution is sensitive to the mass of the electron neutrino. Attempts to measure the endpoint shape of tritium decay have so far seen no distortion from the zero-mass form, thus placing an upper limit of m_nu_beta < 2.3 eV. Here we show that a new type of electron energy spectroscopy could improve future measurements of this spectrum and therefore of the neutrino mass. We propose to detect the coherent cyclotron radiation emitted by an energetic electron in a magnetic field. For mildly relativistic electrons, like those in tritium decay, the relativistic shift of the cyclotron frequency allows us to extract the electron energy from the emitted radiation. We present calculations for the energy resolution, noise limits, high-rate measurement capability, and systematic errors expected in such an experiment.

  5. Facility Microgrids

    SciTech Connect (OSTI)

    Ye, Z.; Walling, R.; Miller, N.; Du, P.; Nelson, K.

    2005-05-01T23:59:59.000Z

    Microgrids are receiving a considerable interest from the power industry, partly because their business and technical structure shows promise as a means of taking full advantage of distributed generation. This report investigates three issues associated with facility microgrids: (1) Multiple-distributed generation facility microgrids' unintentional islanding protection, (2) Facility microgrids' response to bulk grid disturbances, and (3) Facility microgrids' intentional islanding.

  6. Safety System Oversight Assessment of the Los Alamos National Laboratory Weapons Engineering Tritium Facility Tritium Gas Handling System

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011 Strategic2 OPAM615_CostNSARDevelopmental AssignmentAprilANDSafety Software

  7. Evaluation of a tritium runway-lighting system. Technical note

    SciTech Connect (OSTI)

    Katz, E.S.

    1992-04-01T23:59:59.000Z

    A tritium powered runway lighting system was installed and evaluated at the Federal Aviation Administration (FAA) Technical Center. The purpose of this evaluation was to determine if the tritium runway lighting system would safely support Federal Aviation Regulations (FAR) Part 135 commercial operations, during nighttime visual flight rules (VFR) conditions at remote airports. Subject pilots having flight experience levels appropriate for pilots conducting FAR Part 135 air taxi operations were afforded the opportunity of flight testing the system. Results of the evaluation indicate that the tritium runway lighting system does not meet all of the minimum criteria necessary for FAA approval and, therefore, would not guarantee an acceptable level of safety. Tritium Runway Lighting System, Remote Airports.

  8. acute tritium releases: Topics by E-print Network

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

    SAND2012-7340P unknown authors 2012-01-01 34 Revised per Referee's comments October 20, 1999 HEAT DEPOSITION, DAMAGE, AND TRITIUM BREEDING Plasma Physics and Fusion Websites...

  9. Assessment of tritium in the Savannah River Site environment

    SciTech Connect (OSTI)

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

    1993-10-01T23:59:59.000Z

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

  10. Plasma Fueling, Pumping, and Tritium Handling Considerations for FIRE

    SciTech Connect (OSTI)

    Fisher, P.W.; Foster, C.A.; Gentile, C.A.; Gouge, M.J.; Nelson, B.E.

    1999-11-13T23:59:59.000Z

    Tritium pellet injection will be utilized on the Fusion Ignition Research Experiment (FIRE) for efficient tritium fueling and to optimize the density profile for high fusion power. Conventional pneumatic pellet injectors, coupled with a guidetube system to launch pellets into the plasma from the high, field side, low field side, and vertically, will be provided for fueling along with gas puffing for plasma edge density control. About 0.1 g of tritium must be injected during each 10-s pulse. The tritium and deuterium will be exhausted into the divertor. The double null divertor will have 16 cryogenic pumps located near the divertor chamber to provide the required high pumping speed of 200 torr-L/s.

  11. Long Term Tritium Trapping in TFTR and JET

    SciTech Connect (OSTI)

    C.H. Skinner; C.A. Gentile; K.M. Young; J.P. Coad; J.T. Hogan; R.-D. Penzhorn; and N. Bekris

    2001-07-24T23:59:59.000Z

    Tritium retention in TFTR [Tokamak Fusion Test Reactor] and JET [Joint European Torus] shows striking similarities and contrasts. In TFTR, 5 g of tritium were injected into circular plasmas over a 3.5 year period, mostly by neutral-beam injection. In JET, 35 g were injected into divertor plasmas over a 6 month campaign, mostly by gas puffing. In TFTR, the bumper limiter provided a large source of eroded carbon and a major part of tritium was co-deposited on the limiter and vessel wall. Only a small area of the co-deposit flaked off. In JET, the wall is a net erosion area, and co-deposition occurs principally in shadowed parts of the inner divertor, with heavy flaking. In both machines, the initial tritium retention, after a change from deuterium [D] to tritium [T] gas puffing, is high and is due to isotope exchange with deuterium on plasma-facing surfaces (dynamic inventory). The contribution of co-deposition is lower but cumulative, and is revealed by including periods of D fueling that reversed the T/D isotope exchange. Ion beam analysis of flakes from TFTR showed an atomic D/C ratio of 0.13 on the plasma facing surface, 0.25 on the back surface and 0.11 in the bulk. Data from a JET divertor tile showed a larger D/C ratio with 46% C, 30% D, 20% H and 4% O. Deuterium, tritium, and beryllium profiles have been measured and show a thin less than 50 micron co-deposited layer. Flakes retrieved from the JET vacuum vessel exhibited a high tritium release rate of 2e10 Bq/month/g. BBQ modeling of the effect of lithium on retention in TFTR showed overlapping lithium and tritium implantation and a 1.3x increase in local T retention.

  12. Progress in tritium retention and release modeling for ceramic breeders

    SciTech Connect (OSTI)

    Raffray, A.R.; Federici, G. [ITER Joint Work Site, Garching (Germany); Billone, M.C. [Argonne National Lab., IL (United States); Tanaka, S. [Univ. of Tokyo (Japan). Dept. of Quantum Engineering and Systems Science

    1994-07-11T23:59:59.000Z

    Tritium behavior in ceramic breeder blankets is a key design issue for this class of blanket because of its impact on safety and fuel self-sufficiency. Over the past 10-15 years, substantial theoretical and experimental efforts have been dedicated world-wide to develop a better understanding of tritium transport in ceramic breeders. Models that are available today seem to cover reasonably well all the key physical transport and trapping mechanisms. They have allowed for reasonable interpretation and reproduction of experimental data and have helped in pointing out deficiencies in material property data base, in providing guidance for future experiments, and in analyzing blanket tritium behavior. This paper highlights the progress in tritium modeling over the last decade. Key tritium transport mechanisms are briefly described along with the more recent and sophisticated models developed to help understand them. Recent experimental data are highlighted and model calibration and validation discussed. Finally, example applications to blanket cases are shown as illustration of progress in the prediction of ceramic breeder blanket tritium inventory.

  13. Diffusion Coefficient of Tritium Through Molten Salt Flibe and Rate of Tritium Leak from Fusion Reactor System

    SciTech Connect (OSTI)

    Fukada, Satoshi [Kyushu University (Japan); Anderl, Robert A. [Idaho National Engineering and Environmental Laboratory (United States); Sagara, Akio [National Institute for Fusion Science (Japan); Nishikawa, Masabumi [Kyushu University (Japan)

    2005-07-15T23:59:59.000Z

    Diffusion coefficients of hydrogen isotopes in Flibe were correlated with making reference to previous relating data of F{sup -} ion self-diffusivity and Flibe viscosity and so on. Rates of tritium permeation through structural materials in a fusion reactor system with Flibe blanket were estimated comparatively under conditions with or without a Flibe permeation barrier. A way to lower the tritium leak rate below a level regulated by law was proposed, and its effectiveness was discussed.

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

    SciTech Connect (OSTI)

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

    2011-02-23T23:59:59.000Z

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

  15. Facility Safety

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

    1996-10-24T23:59:59.000Z

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

  16. Facility Safety

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

    1995-11-16T23:59:59.000Z

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

  17. Development of a Tritium Extruder for ITER Pellet Injection

    SciTech Connect (OSTI)

    M.J. Gouge; P.W. Fisher

    1998-09-01T23:59:59.000Z

    As part of the International Thermonuclear Experimental Reactor (ITER) plasma fueling development program, Oak Ridge National Laboratory (ORNL) has fabricated a pellet injection system to test the mechanical and thermal properties of extruded tritium. Hydrogenic pellets will be used in ITER to sustain the fusion power in the plasma core and may be crucial in reducing first-wall tritium inventories by a process of "isotopic fueling" in which tritium-rich pellets fuel the burning plasma core and deuterium gas fuels the edge. This repeating single-stage pneumatic pellet injector, called the Tritium-Proof-of-Principle Phase II (TPOP-II) Pellet Injector, has a piston-driven mechanical extruder and is designed to extrude and accelerate hydrogenic pellets sized for the ITER device. The TPOP-II program has the following development goals: evaluate the feasibility of extruding tritium and deuterium-tritium (D-T) mixtures for use in future pellet injection systems; determine the mechanical and thermal properties of tritium and D-T extrusions; integrate, test, and evaluate the extruder in a repeating, single-stage light gas gun that is sized for the ITER application (pellet diameter -7 to 8 mm); evaluate options for recycling propellant and extruder exhaust gas; and evaluate operability and reliability of ITER prototypical fueling systems in an environment of significant tritium inventory that requires secondary and room containment systems. In tests with deuterium feed at ORNL, up to 13 pellets per extrusion have been extruded at rates up to 1 Hz and accelerated to speeds of 1.0 to 1.1 km/s, using hydrogen propellant gas at a supply pressure of 65 bar. Initially, deuterium pellets 7.5 mm in diameter and 11 mm in length were produced-the largest cryogenic pellets produced by the fusion program to date. These pellets represent about a 10% density perturbation to ITER. Subsequently, the extruder nozzle was modified to produce pellets that are almost 7.5-mm right circular cylinders. Tritium and D-T pellets have been produced in experiments at the Los Alamos National Laboratory Tritium Systems Test Assembly. About 38 g of tritium have been utilized in the experiment. The tritium was received in eight batches, six from product containers and two from the Isotope Separation System. Two types of runs were made: those in which the material was only extruded and those in which pellets were produced and fired with deuterium propellant. A total of 36 TZ runs and 28 D-T runs have been made. A total of 36 pure tritium runs and 28 D-T mixture runs were made. Extrusion experiments indicate that both T2 and D-T will require higher extrusion forces than D2 by about a factor of two.

  18. Tritium Effects on Fracture Toughness of Stainless Steel Weldments

    SciTech Connect (OSTI)

    MORGAN, MICHAEL; CHAPMAN, G. K.; TOSTEN, M. H.; WEST, S. L.

    2005-05-12T23:59:59.000Z

    The effects of tritium on the fracture toughness properties of Type 304L and Type 21-6-9 stainless steel weldments were measured. Weldments were tritium-charged-and-aged and then tested in order to measure the effect of the increasing decay helium content on toughness. The results were compared to uncharged and hydrogen-charged samples. For unexposed weldments having 8-12 volume percent retained delta ferrite, fracture toughness was higher than base metal toughness. At higher levels of weld ferrite, the fracture toughness decreased to values below that of the base metal. Hydrogen-charged and tritium-charged weldments had lower toughness values than similarly charged base metals and toughness decreased further with increasing weld ferrite content. The effect of decay helium content was inconclusive because of tritium off-gassing losses during handling, storage and testing. Fracture modes were dominated by the dimpled rupture process in unexposed weldments. In hydrogen and tritium-exposed weldments, the fracture modes depended on the weld ferrite content. At high ferrite contents, hydrogen-induced transgranular fracture of the weld ferrite phase was observed.

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

    DOE Patents [OSTI]

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

    1982-08-19T23:59:59.000Z

    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.

  20. Modeling Tritium Transport in PbLi Breeder Blankets Under Steady State

    SciTech Connect (OSTI)

    H. Zhang; A. Ying; M. Abdou; B. Merrill

    2011-08-01T23:59:59.000Z

    Tritium behavior in the breeder/coolant plays a crucial role in keeping the tritium loss under an allowable limit and realizing high tritium recovery efficiency. In this paper, progress toward the development of a comprehensive 3D predictive capability is discussed and presented. The sequence of transport processes leading to tritium release includes diffusion and convection through the PbLi, transfer across the liquid/solid interface, diffusion of atomic tritium through the structure, and dissolution-recombination at the solid/gas interface. Numerical simulation of the coupled individual physics phenomena of tritium transport is performed for DCLL/HCLL type breeder blankets under realistic reactor-like conditions in this paper. Tritium concentration and permeation are presented and the MHD effects are evaluated. Preliminary results shows that the MHD velocity profile has the significant effect in preventing tritium permeation due to the higher convection effects near the wall.

  1. Tritium flow through a non-symmetrical source. Simulation of gas flow through an injection hole

    E-Print Network [OSTI]

    Sharipov, Felix

    Tritium flow through a non-symmetrical source. Simulation of gas flow through an injection hole of source in injection rarefaction parameter µ0 viscosity of tritium at T0 Pa s 2 #12;Ll = 5074.5 Lr = 5007

  2. 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...

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

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

    Winterberg, F.

    2009-10-29T23:59:59.000Z

    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.

  4. 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.

    2009-01-01T23:59:59.000Z

    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

  5. Dynamic simulation of a proposed ITER tritium processing system

    SciTech Connect (OSTI)

    Kuan, W.; Abdou, M.A. [Univ. of California, Los Angeles, CA (United States); Scott W.R. [Los Alamos National Lab., NM (United States)

    1995-10-01T23:59:59.000Z

    Dynamically simulating the fuel cycle in a fusion reactor is crucial to developing a better understanding of the safe and reliable operation of this complex system. In this work, we propose a tritium processing system for ITER`s plasma exhaust. The dynamic simulation of this proposed system is then performed with the TRUFFLES (TRitiUm Fusion Fuel cycLE dynamic Simulation) model. The fuel management, storage, and fueling operations are developed and coupled with previous cryopump and fuel cleanup unit subsystems to fully realize the complete torus exhaust flow cycle. Results show that tritium inventories will vary widely depending upon reactor operation, individual subsystem and unit operation designs. A diverse collection of batch-controlled subsystems with changes in their processing parameters are simulated in this work. In particular, the effects from the fuel management subsystem`s fuel reserve and tank switching times are quantified using sensitivity studies. 6 refs., 10 figs., 2 tabs.

  6. MODELING TRITIUM TRANSPORT IN PBLI BREEDER BLANKETS UNDER STEADY STATE , M. Abdou1

    E-Print Network [OSTI]

    Abdou, Mohamed

    MODELING TRITIUM TRANSPORT IN PBLI BREEDER BLANKETS UNDER STEADY STATE H. Zhang1 , A. Ying1 , M breeder blankets under realistic reactor-like conditions in this paper. Tritium concentration. Tritium behavior in the liquid metal breeder blanket requires a thorough understanding of the sequence

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

    E-Print Network [OSTI]

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

  8. Intercomparison of tritium and noble gases analyses, 3 and derived parameters excess air and recharge temperature

    E-Print Network [OSTI]

    Intercomparison of tritium and noble gases analyses, 3 H/3 He ages and derived parameters excess with the tritium­helium (3 H/3 He) method has become a powerful tool for hydrogeologists. The uncertainty in the inter- comparison for tritium analyses and ten laboratories participated in the noble gas

  9. Separation phenomena in the tritium source and numerical simulations of turbo-molecular pumps

    E-Print Network [OSTI]

    Sharipov, Felix

    Separation phenomena in the tritium source and numerical simulations of turbo-molecular pumps Felix In the previous works [1, 2], the results of numerical calculations of tritium flow from the buffer vessel up to the first vacuum system were reported. Two values of the tritium source temperature were considered, i.e. 27

  10. PPPL-3311, Preprint: August 1998, UC-420, 423 Modeling of Tritium Retention in TFTR*

    E-Print Network [OSTI]

    - 1 - PPPL-3311, Preprint: August 1998, UC-420, 423 Modeling of Tritium Retention in TFTR* C Fusion Test Reactor tritium retention experience is reviewed and the data related to models of plasma surface interactions. Over 3.5 years of TFTR DT operations, approximately 51% of the tritium injected

  11. Detection of tritium sorption on four soil materials Yanguo Teng a,b

    E-Print Network [OSTI]

    Hu, Qinhong "Max"

    Detection of tritium sorption on four soil materials Yanguo Teng a,b , Rui Zuo a,b,*, Jinsheng Wang December 2010 Keywords: Tritium Adsorption Distribution coefficient water/solid ratio pH Humic substances, it is important to understand the sorption behavior of tritium on soils. In this study, batch tests were carried

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

    E-Print Network [OSTI]

    Tritium Reduction and Control in the Vacuum Vessel During TFTR Outage and Decommissioning * W of the torus, a three tier system was developed for the outage in order to reduce and control the free tritium. The first phase of the program to reduce the free tritium consisted of direct flowthrough of room air

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

    E-Print Network [OSTI]

    Tritium Reduction and Control in the Vacuum Vessel During TFTR Outage and Decommissioning* W, a three tier system was developed for the outage in order to reduce and control the free tritium. The first phase of the program to reduce the free tritium consisted of direct flowthrough of room air

  14. APPLICATION OF BIOASSAY FOR TRITIUM A. CONDITIONS UNDER WHICH BIOASSAY IS NECESSARY

    E-Print Network [OSTI]

    Slatton, Clint

    APPENDIX F APPLICATION OF BIOASSAY FOR TRITIUM A. CONDITIONS UNDER WHICH BIOASSAY IS NECESSARY 1. Routine bioassay is necessary when quantities of tritium processed by an individual at any one time or the total amount processed per month exceed those for the forms of tritium shown in Table 1. Table 1

  15. DIFFUSION ELASTIQUE DES NEUTRONS PAR LE TRITIUM A 14 MeV

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    DIFFUSION ELASTIQUE DES NEUTRONS PAR LE TRITIUM A 14 MeV Laboratoire de Neutronique, CNRS, Toulouse of neutrons from tritium is studied with a thin scatterer close to a Cs1 scintillator. The experimental. Le diffuseur utilisé est une cible mince contenant 2,22 an3 de tritium absorbé dans une épaisseur de

  16. Monte Carlo Calculations of the Intrinsic Detector Backgrounds for the Karlsruhe Tritium Neutrino Experiment

    E-Print Network [OSTI]

    Washington at Seattle, University of - Department of Physics, Electroweak Interaction Research Group

    Monte Carlo Calculations of the Intrinsic Detector Backgrounds for the Karlsruhe Tritium Neutrino of the Intrinsic Detector Backgrounds for the Karlsruhe Tritium Neutrino Experiment Michelle L. Leber Chair of the Supervisory Committee: Professor John F. Wilkerson Physics The Karlsruhe Tritium Neutrino Experiment (KATRIN

  17. CALCULATIONS OF TRITIUM FLOW BETWEEN THE BUFFER VESSEL UP TO THE FIRST VACUUM SYSTEM

    E-Print Network [OSTI]

    Sharipov, Felix

    CALCULATIONS OF TRITIUM FLOW BETWEEN THE BUFFER VESSEL UP TO THE FIRST VACUUM SYSTEM Felix Sharipov diff., Eq.(32) µ viscosity of tritium Pa s 1 Introduction The present work is a continuation of the previous report [1], where the preliminary results were obtained for the tritium flow through the source

  18. Tritium Containment in the Dust and Debris of Plasma-Facing Materials Produced During Operations

    E-Print Network [OSTI]

    Harilal, S. S.

    ' . . , . Tritium Containment in the Dust and Debris of Plasma-Facing Materials Produced During avaihble original document. #12;Tritium Containment in the Dust and Debris of Plasma-Facing hlaterials. IL 60439, USA Tritium behavior in plasma-facing components of future tokamak reactors such as ITER

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

    E-Print Network [OSTI]

    PPPL-3458 PPPL-3458 UC-70 Visual Tritium Imaging Of In-Vessel Surfaces by C. A. Gentile, S. J: http://www.ntis.gov/ordering.htm #12;1 Visual Tritium Imaging Of In-Vessel Surfaces C. A. Gentile, S. J Energy Research Institute, Tritium Engineering Laboratory, Tokai, Ibaraki 319-1195, Japan Abstract

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

    SciTech Connect (OSTI)

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

    1983-05-18T23:59:59.000Z

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

  1. Fusion Engineering and Design 81 (2006) 14651470 Influence of 2D and 3D convectiondiffusion flow on tritium

    E-Print Network [OSTI]

    Abdou, Mohamed

    2006-01-01T23:59:59.000Z

    on tritium permeation in helium cooled solid breeder blanket units Wen Guo, Alice Ying, Ming-Jiu Ni, Mohamed; accepted 23 August 2005 Available online 10 January 2006 Abstract Numerical simulation of tritium, and transient diffusion and convection equations are simulated for the tritium permeation analysis. Tritium

  2. TPOP-II: Tritium Fueling at a Re P. W. Fisher and M. J. Gouge Oak Ridge National Labora

    E-Print Network [OSTI]

    TPOP-II: Tritium Fueling at a Re Scale P. W. Fisher and M. J. Gouge Oak Ridge National Labora properties of extruded t repeating single-stage pneumatic pellet injector, called the Tritium-Proof-o Phase at the Los Alamos Nati Laboratory Tritium Systems Test Assembly (TSTA). About 38 g of tritium utilized

  3. Facility Safety

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

    2005-12-22T23:59:59.000Z

    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.

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

    E-Print Network [OSTI]

    Ragusa, Jean Concetto

    1996-01-01T23:59:59.000Z

    investigated. The target designs in the heterogeneous systems were 1 / liquid lead, and 2/ layers of solid lead plates cooled by heavy water. The tritium breeding blanket assemblies contained either lithium oxide or molten fluorine salt with or without UF4...

  5. Test of potential homogeneity in the KATRIN gaseous tritium source

    E-Print Network [OSTI]

    M. Rysavy

    2005-06-02T23:59:59.000Z

    83mKr is supposed to be used to study the properties of the windowless gaseous tritium source of the experiment KATRIN. In this work we deduce the amount of 83mKr which is necessary to determine possible potential inhomogeneities via conversion-electron-line broadening.

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

    SciTech Connect (OSTI)

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

    1983-01-01T23:59:59.000Z

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

  7. Material Sample Collection with Tritium and Gamma Analyses at the University of Illinois's Nuclear Research Laboratory TRIGA Nuclear Research Reactor

    SciTech Connect (OSTI)

    Charters, G.; Aggarwal, S. [New Millennium Nuclear Technologies, 575 Union Blvd, Suite 102, Lakewood, CO 80228 (United States)

    2006-07-01T23:59:59.000Z

    The University of Illinois in Champaign-Urbana has an Advanced TRIGA reactor facility which was built in 1960 and operated until August 1998. The facility was shutdown for a variety of reasons, primarily due to a lack of usage by the host institution. In 1998 the reactor went into SAFSTOR and finally shipped its fuel in 2004. At the present time a site characterization and decommissioning plan are in process and hope to be submitted to the NRC in early 2006. The facility had to be fully characterized and part of this characterization involved the collection and analysis of samples. This included various solid media such as, concrete, graphite, metals, and sub-slab surface soils for immediate analysis of Activation and Tritium contamination well below the easily measured surfaces. This detailed facility investigation provided a case to eliminate historical unknowns, increasing the confidence for the segregation and packaging of high specific activity Low Level Radwaste (LLRW), from which a strategy of 'surgical-demolition' and segregation could be derived thus maximizing the volumes of 'clean material'. Performing quantitative volumetric concrete or metal radio-analyses safer and faster (without lab intervention) was a key objective of this dynamic characterization approach. Currently, concrete core bores are shipped to certified laboratories where the concrete residue is run through a battery of tests to determine the contaminants. The existing core boring operation volatilises or washes out some of the contaminants (like tritium) and oftentimes cross-contaminates the are a around the core bore site. The volatilization of the contaminants can lead to airborne problems in the immediate vicinity of the core bore. Cross-contamination can increase the contamination area and thereby increase the amount of waste generated that needs to be treated and stabilized before disposal. The goal was to avoid those field activities that could cause this type of release. Therefore, TRUPRO{sup R}, a sampling and profiling tool in conjunction with radiometric instrumentation was utilized to produce contamination profiles through the material being studied. All samples (except metals) on-site were analyzed within 10 minutes for tritium using a calibrated portable liquid scintillation counter (LSC) and analyzed for gamma activation products using a calibrated ISOCS. Improved sample collection with near real time analysis along with more historical hazard analysis enhanced significantly over the baseline coring approach the understanding of the depth distribution of contaminants. The water used in traditional coring can result in a radioactive liquid waste that needs to be dealt with. This would have been an issue at University of Illinois. Considerable time, risk reduction and money are saved using this profiling approach. (authors)

  8. Topical report on a preconceptual design for the Spallation-Induced Lithium Conversion (SILC) target for the accelerator production of tritium (APT)

    SciTech Connect (OSTI)

    Van Tuyle, G.J.; Cokinos, D.M.; Czajkowski, C.; Franz, E.M.; Kroeger, P.; Todosow, M.; Youngblood, R.; Zucker, M.

    1993-09-30T23:59:59.000Z

    The preconceptual design of the APT Li-Al target system, also referred to as the Spallation-Induced Lithium Conversion (SILC), target system, is summarized in this report. The system has been designed to produce a ``3/8 Goal`` quantity of tritium using the 200-mA, 1.0 GeV proton beam emerging from the LANL-designed LINAC. The SILC target system consists of a beam expander, a heavy-water-cooled lead spallation neutron source assembly surrounded by light-water-cooled Li-Al blankets, a target window, heat removal systems, and related safety systems. The preconceptual design of each of these major components is described. Descriptions are also provided for the target fabrication, tritium extraction, and waste-steam processes. Performance characteristics are presented and discussed.

  9. Facility Safety

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

    2002-05-20T23:59:59.000Z

    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. Treatment of gaseous effluents at nuclear facilities

    SciTech Connect (OSTI)

    Goossen, W.R.A. [Studiecentrum voor Kernenergie, Mol (Belgium). Dept. of Chemical Engineering] [ed.; Eichholz, G.G.; Tedder, D.W. [eds.] [Georgia Institute of Technology, Atlanta, GA (United States)

    1991-12-31T23:59:59.000Z

    Airborne effluents from nuclear facilities represent the major environmental impact from such plants both under routine conditions or after plant accidents. Effective control of such emissions, therefore, constitutes a major aspect of plant design for nuclear power plants and other facilities in the nuclear fuel cycle. This volume brings together a number of review articles by experts in the various areas of concern and describes some of the removal systems that have been designed for power plants and, particularly, for reprocessing plants. Besides controlling the release of radionuclides, other potentially hazardous effluents, such as nitrous oxides, must be minimized, and these are included in some of the systems described. The various chapters deal with historic developments and current technology in reducing emission of fission products, noble gases, iodine, and tritium, and consider design requirements for practical installations.

  11. Facility Safety

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

    2005-12-22T23:59:59.000Z

    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.

  12. Facility Safety

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

    2013-06-21T23:59:59.000Z

    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.

  13. Facility Safety

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

    2000-11-20T23:59:59.000Z

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

  14. FINITE ELEMENT ANALYSIS OF BULK TRITIUM SHIPPING PACKAGE

    SciTech Connect (OSTI)

    Jordan, J.

    2010-06-02T23:59:59.000Z

    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.

  15. Results from deuterium-tritium tokamak confinement experiments

    SciTech Connect (OSTI)

    Hawryluk, R.J.

    1997-02-01T23:59:59.000Z

    Recent scientific and technical progress in magnetic fusion experiments has resulted in the achievement of plasma parameters (density and temperature) which enabled the production of significant bursts of fusion power from deuterium-tritium fuels and the first studies of the physics of burning plasmas. The key scientific issues in the reacting plasma core are plasma confinement, magnetohydrodynamic (MHD) stability, and the confinement and loss of energetic fusion products from the reacting fuel ions. Progress in the development of regimes of operation which have both good confinement and are MHD stable have enabled a broad study of burning plasma physics issues. A review of the technical and scientific results from the deuterium-tritium experiments on the Joint European Torus (JET) and the Tokamak Fusion Test Reactor (TFTR) is given with particular emphasis on alpha-particle physics issues.

  16. DEVELOPMENT AND USE OF A BULK TRITIUM SHIPPING PACKAGE

    SciTech Connect (OSTI)

    Blanton, P.

    2010-09-30T23:59:59.000Z

    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.

  17. Direct determination of Neutrino Mass from Tritium Beta Spectrum

    E-Print Network [OSTI]

    C. Weinheimer

    2009-12-08T23:59:59.000Z

    The investigation of the endpoint region of the tritium beta decay spectrum is still the most sensitive direct method to determine the neutrino mass scale. In the nineties and the beginning of this century the tritium beta decay experiments at Mainz and Troitsk have reached a sensitivity on the neutrino mass of 2 eV/c^2 . They were using a new type of high-resolution spectrometer with large sensitivity, the MAC-E-Filter, and were studying the systematics in detail. Currently, the KATRIN experiment is being set up at Forschungszentrum Karlsruhe, Germany. KATRIN will improve the neutrino mass sensitivity by one order of magnitude down to 0.2 eV/c^2, sufficient to cover the degenerate neutrino mass scenarios and the cosmologically relevant neutrino mass range.

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

    SciTech Connect (OSTI)

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

    2013-10-02T23:59:59.000Z

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

  19. Predicted versus measured tritium oxide concentrations at SRS

    SciTech Connect (OSTI)

    Simpkins, A.A.

    1995-12-01T23:59:59.000Z

    Measured tritium oxide concentrations at various offsite locations are compared with concentrations predicted by three computer codes that are utilized at SRS to predict doses to the maximally exposed offsite individuals. Annual average concentrations predicted by the computer programs were compared with measured average concentrations taken form data contained in the last ten years of SRS Environmental Reports. The computer programs used for the comparison are ACAIRQ, CAP88, and MAXIGASP.

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

    DOE Patents [OSTI]

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

    1989-12-26T23:59:59.000Z

    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.

  1. 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)

    1989-01-01T23:59:59.000Z

    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.

  2. CIBLE DE TRITIUM LIQUIDE M. CHEMARIN, L. FEUVRAIS, M. GOUANRE, M. C. LEMAIRE et G. NICOLA,

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    236. CIBLE DE TRITIUM LIQUIDE M. CHEMARIN, L. FEUVRAIS, M. GOUAN�RE, M. C. LEMAIRE et G. NICOLAÏ, Institut de Physique Nucléaire, Lyon. Résumé. - On présente une cible de tritium ou deutérium liquide de petite dimension : 0,4 cm3 de tritium liquide, soit une perte d'énergie de 1,4 MeV pour des deutons de 27

  3. EIS-0270: Accelerator Production of Tritium at the Savannah River Site

    Broader source: Energy.gov [DOE]

     This EIS evaluates the potential environmental impact of a proposal to construct and operate an Accelerator for the Production of Tritium at the Savannah River Site.  

  4. An experience of use of the installation for the cleaning of gas effluents from tritium

    SciTech Connect (OSTI)

    Voitenko, V.A.; Kolomiets, N.F.; Rogosin, V.N.

    1993-12-31T23:59:59.000Z

    The population and environmental protection during the operation of nuclear engineering units is a serious scientific-technical and social problem. Tritium is one of the gaseous effluents from nuclear plants, reactor fuel element processing, and also in connection with perspective thermo-nuclear power engineering development. The authors propose the use of a cleaning system for gas effluent cleaning of tritium using catalysis methods. The process of catalytic gas cleaning involves chemical transformations resulting in the removal of impurities from the reaction mixture. The technological equipment for tritium treatment is intended for production of such items on tritium bases as neutron tubes, targets, sources of initial ionization and characteristic rays, etc.

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

    DOE Patents [OSTI]

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

    2001-01-01T23:59:59.000Z

    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.

  6. Facility Safety

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

    1995-10-13T23:59:59.000Z

    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.

  7. Facility Safety

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

    2012-12-04T23:59:59.000Z

    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.

  8. Type A Investigation - Subcontractor Fatality at the Savannah...

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

    2, 2002, Worker Fall from ShoringScaffolding Structure at the Savannah River Site Tritium Extraction Facility Construction Site Type A Accident Investigation Board Report on...

  9. Type B Investigation Board Report Worker Fall from Shoring/Scaffolding...

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

    Scaffolding Structure at the Savannah River Site Tritium Extraction Facility - Construction Site April 2, 2002 SAVANNAH RIVER SITE OFFICE MAY 2002 Worker Fall from Shoring...

  10. 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.

    2005-08-30T23:59:59.000Z

    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.

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

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

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

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742EnergyOn April 23, 2014, an OHASeptember 2010In addition to 1 |DDOE HDBK-1113-2008

  13. VAPOR PRESSURE ISOTOPE EFFECTS IN THE MEASUREMENT OF ENVIRONMENTAL TRITIUM SAMPLES.

    SciTech Connect (OSTI)

    Kuhne, W.

    2012-12-03T23:59:59.000Z

    Standard procedures for the measurement of tritium in water samples often require distillation of an appropriate sample aliquot. This distillation process may result in a fractionation of tritiated water and regular light water due to the vapor pressure isotope effect, introducing either a bias or an additional contribution to the total tritium measurement uncertainty. The magnitude of the vapor pressure isotope effect is characterized as functions of the amount of water distilled from the sample aliquot and the heat settings for the distillation process. The tritium concentration in the distillate is higher than the tritium concentration in the sample early in the distillation process, it then sharply decreases due to the vapor pressure isotope effect and becomes lower than the tritium concentration in the sample, until the high tritium concentration retained in the boiling flask is evaporated at the end of the process. At that time, the tritium concentration in the distillate again overestimates the sample tritium concentration. The vapor pressure isotope effect is more pronounced the slower the evaporation and distillation process is conducted; a lower heat setting during the evaporation of the sample results in a larger bias in the tritium measurement. The experimental setup used and the fact that the current study allowed for an investigation of the relative change in vapor pressure isotope effect in the course of the distillation process distinguish it from and extend previously published measurements. The separation factor as a quantitative measure of the vapor pressure isotope effect is found to assume values of 1.034 {+-} 0.033, 1.052 {+-} 0.025, and 1.066 {+-} 0.037, depending on the vigor of the boiling process during distillation of the sample. A lower heat setting in the experimental setup, and therefore a less vigorous boiling process, results in a larger value for the separation factor. For a tritium measurement in water samples, this implies that the tritium concentration could be underestimated by 3 - 6%.

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

    SciTech Connect (OSTI)

    Ivey, Wade

    2013-12-17T23:59:59.000Z

    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.

  15. Dual Feedback Controlled High Performance Ar Seeded ELMy H-mode Discharges in JET including Trace Tritium Experiments

    E-Print Network [OSTI]

    Dual Feedback Controlled High Performance Ar Seeded ELMy H-mode Discharges in JET including Trace Tritium Experiments

  16. The Analytical Gas Chromatographic System of the JET Active Gas Handling System ­ Tritium Commissioning and use during DTE1

    E-Print Network [OSTI]

    The Analytical Gas Chromatographic System of the JET Active Gas Handling System ­ Tritium Commissioning and use during DTE1

  17. Evidence of a Pathway to Hydrocarbon Nanoparticle Formation in Fusion Plasmas and its Impact on Tritium Inventory

    E-Print Network [OSTI]

    Evidence of a Pathway to Hydrocarbon Nanoparticle Formation in Fusion Plasmas and its Impact on Tritium Inventory

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

    E-Print Network [OSTI]

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

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

    E-Print Network [OSTI]

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

  20. Determination of Tritium Profiles in Tiles from the First Wall of Fusion Machines and Development of Techniques for their Detritiation

    E-Print Network [OSTI]

    Determination of Tritium Profiles in Tiles from the First Wall of Fusion Machines and Development of Techniques for their Detritiation

  1. Transport Analysis of Trace Tritium Experiments on JET using TRANSP Code and Comparison with Theory-Based Transport Models

    E-Print Network [OSTI]

    Transport Analysis of Trace Tritium Experiments on JET using TRANSP Code and Comparison with Theory-Based Transport Models

  2. A New Interpretation of Alpha-Particle-Driven Instabilities in Deuterium-Tritium Experiments on the Tokamak Fusion Test Reactor

    E-Print Network [OSTI]

    A New Interpretation of Alpha-Particle-Driven Instabilities in Deuterium-Tritium Experiments on the Tokamak Fusion Test Reactor

  3. Tritium Behavior in Eroded Dust and Debris of Plasma-Facing Materials*

    E-Print Network [OSTI]

    Harilal, S. S.

    . & ,. 1. . . . Tritium Behavior in Eroded Dust and Debris of Plasma-Facing Materials* RECQVED SEP2 Reactor Materials (ICFRM-8) October 26-31, 1997, Sendai, Japan. qWork supported by the U.S. Department;. . . . . Tritium Behavior in Eroded Dust and Debris of Plasma-Facing Materials A. Hassanein', B. Wiechers2, and I

  4. TRITIUM RECYCLING AND INVENTORY IN ERODED DEBRIS OF PLASMA-FACING MATERIALS*

    E-Print Network [OSTI]

    Harilal, S. S.

    .,, TRITIUM RECYCLING AND INVENTORY IN ERODED DEBRIS OF PLASMA-FACING MATERIALS* Ahmed Hassanein RECYCLING AND INVENTORY IN ERODED DEBRIS OF PLASMA-FACING MATERIALS AmvlED H.ASSANEIN Argonne Mm therefore, they can significantly infIuence plasma behavior and tritium inventory during subsequent

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

    DOE Patents [OSTI]

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

    1984-10-09T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    2001-11-12T23:59:59.000Z

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

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

    DOE Patents [OSTI]

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

    1984-10-09T23:59:59.000Z

    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.

  8. Characteristics of the Yala Glacier from theview point of tritium content

    SciTech Connect (OSTI)

    Kamiyama, K.; Kitaoka, K.; Watanabe, O.

    1986-10-20T23:59:59.000Z

    The vertical distributions of tritium content in glacier ice were determined at two different heights of the Yala Glacier in the Langtang Region, Nepal Himalayas. In the vertical profile of tritium content at the upper point (about 5,400 m high) there exists a clear peak, which shows the injection of artificial tritium due to nuclear weapon test series. The average accumulation rate from 1963 to 1982 at this point is estimated to be 0.85 m of water equivalent per year. At the lower point (about 5,200 m high) the tritium content is relatively low throughout the ice core, decreasing with depth. Precipitation nourishes the glacier in the accumulation area and, after a long time, it appears in the ablation area with the movement of the ice body, resulting in the different profiles of the tritium content between the two points. From the viewpoint of tritium content, the precipitation in the Yala Glacier is more similar to that in New Delhi, India, than that in Karizimir, Afghanistan. Generally speaking, the tritium content in the pricipitation is lower in the coastal area than in the continental area. The precipitation in the Yala Glacier belongs to the coastal type. There possibly exists a great difference in tritium content between the glaciers nourished by water vapor coming directly from the sea and by that coming over the continent.

  9. EFFECTS OF ONE WEEK TRITIUM EXPOSURE ON EPDM ELASTOMER

    SciTech Connect (OSTI)

    Clark, E

    2007-06-07T23:59:59.000Z

    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.

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

    DOE Patents [OSTI]

    Heung, Leung K. (Aiken, SC)

    1994-01-01T23:59:59.000Z

    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.

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

    DOE Patents [OSTI]

    Heung, L.K.

    1994-03-29T23:59:59.000Z

    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.

  12. Tritium Instrument Demonstration Station (TIDS) | Department of Energy

    Office of Environmental Management (EM)

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

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011 Strategic2Uranium Transferon theTedRegion | Department of Energy4th Tritium

  14. Brookhaven National Laboratory - HFBR Tritium | Department of Energy

    Office of Environmental Management (EM)

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  15. Monitoring of Tritium release at PTC | Department of Energy

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011 Strategic2 OPAM615_Cost Estimating35.doc Microsoft2006 |VehicleofMonitoring of Tritium

  16. PHYSICAL REVIEW C VOLUME 27, NUMBER 4 APRIL 1983 Atomic final-state interactions in tritium decay

    E-Print Network [OSTI]

    Williams, Roy

    PHYSICAL REVIEW C VOLUME 27, NUMBER 4 APRIL 1983 Atomic final-state interactions in tritium decay R of the ejected P ray with the bound atomic e1ectron in the P decay of a tritium atom. The excited state probabi1 effects are expected to be more pro- nounced, but not, to our knowledge, for tritium. The interaction

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

    E-Print Network [OSTI]

    Aeschbach-Hertig, Werner

    Physics of Aquatic Systems II, 6. Tritium Universität HeidelbergInstitut für Umweltphysik 1 Physics of Aquatic Systems II ­ 6. Tritium Werner Aeschbach-Hertig Institute of Environmental Physics University of Heidelberg Physics of Aquatic Systems II, 6. Tritium Universität HeidelbergInstitut für Umweltphysik 2

  18. Trace tritium and the H-mode density limit G.F. Matthews a,*, K.-D. Zastrow a

    E-Print Network [OSTI]

    Basse, Nils Plesner

    Trace tritium and the H-mode density limit G.F. Matthews a,*, K.-D. Zastrow a , P. Andrew a , B University, Princeton, NJ 08543, USA Abstract Trace amounts of tritium gas have been injected in short pus then saturation of the separatrix density implies clamping of the core density. Short pus of tritium were used

  19. Tritium behavior in eroded dust and debris of plasma-facing A. Hassanein a,*, B. Wiechers b

    E-Print Network [OSTI]

    Harilal, S. S.

    Tritium behavior in eroded dust and debris of plasma-facing materials A. Hassanein a,*, B. Wiechers, Russian Federation Abstract Tritium behavior in plasma-facing components (PFCs) of future tokamak reactors important parameter that inŻuences tritium buildup and release in candidate materials is the eect

  20. Investigation of plasma-facing component material effects on tritium reprocessing systems

    SciTech Connect (OSTI)

    Kuan, W.; Abdou, M.A. [Univ. of California, Los Angeles, CA (United States)

    1995-10-01T23:59:59.000Z

    Plasma-facing component (PFC) materials directly affect tritium inventories by the creation of a characteristic set of volatile impurities inside the torus. Impurity creation processes were modeled and incorporated into the TritiUm Fusion Fuel cycLE dynamic Simulation, TRUFFLES, which simulates dynamic inventories in the tritium reprocessing systems. These surface processes include net erosion and `outgassing`. The estimated impurity outflow is coupled with the tritium reprocessing models in TRUFFLES to calculate inventories. Be and C were evaluated as examples of plasma-facing materials. It is found that for C a constraint limiting its net erosion rate is necessary in order to keep the tritium inventory in the cryopumps below a specified value. In contrast, Be may present no problem because of its non-production of volatile species when eroded during reactor power operation. `Outgassing` of H{sub 2}O and the DT reflection coefficient were also investigated. 8 refs., 5 figs., 2 tabs.

  1. Fluid extraction

    DOE Patents [OSTI]

    Wai, Chien M. (Moscow, ID); Laintz, Kenneth E. (Los Alamos, NM)

    1999-01-01T23:59:59.000Z

    A method of extracting metalloid and metal species from a solid or liquid material by exposing the material to a supercritical fluid solvent containing a chelating agent is described. The chelating agent forms chelates that are soluble in the supercritical fluid to allow removal of the species from the material. In preferred embodiments, the extraction solvent is supercritical carbon dioxide and the chelating agent is a fluorinated .beta.-diketone. In especially preferred embodiments the extraction solvent is supercritical carbon dioxide, and the chelating agent comprises a fluorinated .beta.-diketone and a trialkyl phosphate, or a fluorinated .beta.-diketone and a trialkylphosphine oxide. Although a trialkyl phosphate can extract lanthanides and actinides from acidic solutions, a binary mixture comprising a fluorinated .beta.-diketone and a trialkyl phosphate or a trialkylphosphine oxide tends to enhance the extraction efficiencies for actinides and lanthanides. The method provides an environmentally benign process for removing contaminants from industrial waste without using acids or biologically harmful solvents. The method is particularly useful for extracting actinides and lanthanides from acidic solutions. The chelate and supercritical fluid can be regenerated, and the contaminant species recovered, to provide an economic, efficient process.

  2. Mobile Facility

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

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  3. Facility Representatives

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011AT&T,OfficeEnd ofEvaluations in Covered Facilities | Department of Energy

  4. Facility Representatives

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011AT&T,OfficeEnd ofEvaluations in Covered Facilities | Department of Energy063-2011

  5. Facility Status

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOEThe Bonneville Power AdministrationField8,Dist. Category UC-lFederalFYRANDOM DRUG TESTING The requirementFacility

  6. Accelerator Production of Tritium project process waste assessment

    SciTech Connect (OSTI)

    Carson, S.D.; Peterson, P.K.

    1995-09-01T23:59:59.000Z

    DOE has made a commitment to compliance with all applicable environmental regulatory requirements. In this respect, it is important to consider and design all tritium supply alternatives so that they can comply with these requirements. The management of waste is an integral part of this activity and it is therefore necessary to estimate the quantities and specific wastes that will be generated by all tritium supply alternatives. A thorough assessment of waste streams includes waste characterization, quantification, and the identification of treatment and disposal options. The waste assessment for APT has been covered in two reports. The first report was a process waste assessment (PWA) that identified and quantified waste streams associated with both target designs and fulfilled the requirements of APT Work Breakdown Structure (WBS) Item 5.5.2.1. This second report is an expanded version of the first that includes all of the data of the first report, plus an assessment of treatment and disposal options for each waste stream identified in the initial report. The latter information was initially planned to be issued as a separate Waste Treatment and Disposal Options Assessment Report (WBS Item 5.5.2.2).

  7. PUREX facility hazards assessment

    SciTech Connect (OSTI)

    Sutton, L.N.

    1994-09-23T23:59:59.000Z

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

  8. Extractant composition

    DOE Patents [OSTI]

    Smith, Barbara F. (Los Alamos, NM); Jarvinen, Gordon D. (Los Alamos, NM); Ryan, Robert R. (Los Alamos, NM)

    1990-01-01T23:59:59.000Z

    An organic extracting solution useful for separating elements of the actinide series of the periodic table from elements of the lanthanide series, where both are in trivalent form. The extracting solution consists of a primary ligand and a secondary ligand, preferably in an organic solvent. The primary ligand is a substituted monothio-1,3-dicarbonyl, which includes a substituted 4-acyl-2-pyrazolin-5-thione, such as 4-benzoyl-2,4-dihydro-5-methyl-2-phenyl-3H-pyrazol-3-thione (BMPPT). The secondary ligand is a substituted phosphine oxide, such as trioctylphosphine oxide (TOPO).

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

    E-Print Network [OSTI]

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

    2013-06-03T23:59:59.000Z

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

  10. The data collection system for failure/maintenance at the Tritium Systems Test Assembly

    SciTech Connect (OSTI)

    Casey, M.A.; Gruetzmacher, K.M.; Bartlit, J.R.; Cadwallader, L.C.

    1988-01-01T23:59:59.000Z

    A data collection system for obtaining information which can be used to help determine the reliability and vailability of future fusion power plants has been installed at the Los Alamos National Laboratory's Tritium Systems Test Assembly (TSTA). Failure and maintenance data on components of TSTA's tritium systems have been collected since 1984. The focus of the data collection has been TSTA's Tritium Waste Tratment System (TWT), which has maintained high availability since it became operation in 1982. Data collection is still in progress and a total of 291 failure reports are in the data collection system at this time, 47 of which are from the TWT. 6 refs., 2 figs., 2 tabs.

  11. Determination of tritium activity in environmental water samples using gas analyzer techniques

    E-Print Network [OSTI]

    Salsman, John Matthew

    1983-01-01T23:59:59.000Z

    undesirable when looking for low levels of tritium activity in water. In general, gas analyzer techniques consist of dispersing the radioactive material in some type of gaseous medium and then pressurizing the system with this gas. The analyzer then uses... by reaction with gaseous hydrogen. The vapor HTO is formed readily, as shown by. Equation 2, and is the most commonly encountered form of tritium in the environment. HT + 820 H2 + HTO (2) The accumulation of tritium on the Earth occurs both naturally...

  12. Environmental and Economical Evaluation of Integrating NGL Extraction and LNG Liquefaction Technology in Iran LNG Project 

    E-Print Network [OSTI]

    Manesh, M. H. K.; Mazhari, V.

    2009-01-01T23:59:59.000Z

    LNG and NGL for comparable compression schemes as compared to stand-alone LNG liquefaction and NGL extraction facilities. In addition, there are potential enhancements to the overall facility availability and project economics and environmental impacts...

  13. Hydrodynamic instability growth and mix experiments at the National Ignition Facility

    SciTech Connect (OSTI)

    Smalyuk, V. A.; Barrios, M.; Caggiano, J. A.; Casey, D. T.; Cerjan, C. J.; Clark, D. S.; Edwards, M. J.; Haan, S. W.; Hammel, B. A.; Hamza, A.; Hsing, W. W.; Hurricane, O.; Kroll, J.; Landen, O. L.; Lindl, J. D.; Ma, T.; McNaney, J. M.; Mintz, M.; Parham, T.; Peterson, J. L. [Lawrence Livermore National Laboratory, NIF Directorate, Livermore, California 94550 (United States)] [Lawrence Livermore National Laboratory, NIF Directorate, Livermore, California 94550 (United States); and others

    2014-05-15T23:59:59.000Z

    Hydrodynamic instability growth and its effects on implosion performance were studied at the National Ignition Facility [G. H. Miller, E. I. Moses, and C. R. Wuest, Opt. Eng. 443, 2841 (2004)]. Implosion performance and mix have been measured at peak compression using plastic shells filled with tritium gas and containing embedded localized carbon-deuterium diagnostic layers in various locations in the ablator. Neutron yield and ion temperature of the deuterium-tritium fusion reactions were used as a measure of shell-gas mix, while neutron yield of the tritium-tritium fusion reaction was used as a measure of implosion performance. The results have indicated that the low-mode hydrodynamic instabilities due to surface roughness were the primary culprits for yield degradation, with atomic ablator-gas mix playing a secondary role. In addition, spherical shells with pre-imposed 2D modulations were used to measure instability growth in the acceleration phase of the implosions. The capsules were imploded using ignition-relevant laser pulses, and ablation-front modulation growth was measured using x-ray radiography for a shell convergence ratio of ?2. The measured growth was in good agreement with that predicted, thus validating simulations for the fastest growing modulations with mode numbers up to 90 in the acceleration phase. Future experiments will be focused on measurements at higher convergence, higher-mode number modulations, and growth occurring during the deceleration phase.

  14. Metal extraction

    SciTech Connect (OSTI)

    Covington, J.W.; Whittemore, R.G.

    1980-10-21T23:59:59.000Z

    In a process according to the present invention uranium is extracted into solution from its ore by leaching with an aqueous solution containing peroxomonosulphuric acid, the peroxoacid oxidizing the uranium through to its hexavalent state. Preferably the leaching is carried out at a temperature in the range of 50* to 100* C. The leach liquor can initially contain additional amounts of sulphuric acid or merely that present by virtue of the method of making the peroxomonosulphuric acid. In a preferred method of operation, the peroxoacid is introduced progressively into the leach liquor during the course of the leaching so as to maintain an electrochemical potential in the range of 450 to 650 mV. By use of the process, uranium is cleanly extracted into solution.

  15. Measurement of limiter heating due to fusion product losses during high fusion power deuterium-tritium operation of TFTR

    SciTech Connect (OSTI)

    Janos, A.; Owens, D.K.; Darrow, D.; Redi, M.; Zarnstorff, M.; Zweben, S.

    1995-03-01T23:59:59.000Z

    Preliminary analysis has been completed on measurements of limiter heating during high fusion power deuterium-tritium (D-T) operation of TFTR, in an attempt to identify heating from alpha particle losses. Recent operation of TFTR with a 50-50 mix of D-T has resulted in fusion power output ({approx} 6.2 MW) orders of magnitude above what was previously achieved on TFTR. A significantly larger absolute number of particles and energy from fusion products compared to D-D operation is expected to be lost to the limiters. Measurements were made in the vicinity of the midplane ({plus_minus} 30{degree}) with thermocouples mounted on the tiles of an outboard limiter. Comparisons were made -between discharges which were similar except for the mix of deuterium and tritium beam sources. Power and energy estimates of predicted alpha losses were as high as 0.13 MW and 64 kJ. Depending on what portion of the limiters absorbed this energy, temperature rises of up to 42 {degrees}C could be expected, corresponding to a heat load of 0.69 MJ/m{sup 2} over a 0.5 sec period, or a power load of 1.4 MW/m{sup 2}. There was a measurable increase in the limiter tile temperature as the fusion power yield increased with a more reactive mixture of D and T at constant beam power during high power D-T operation. Analysis of the data is being conducted to see if the alpha heating component can be extracted. Measured temperature increases were no greater than 1 {degree}C, indicating that there was probably neither an unexpectedly large fraction of lost particles nor unexpected localization of the losses. Limits on the stochastic ripple loss contribution from alphas can be deduced.

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

    SciTech Connect (OSTI)

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

    2011-05-01T23:59:59.000Z

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

  17. Tritium processing system for the ITER Li/V blanket test module

    SciTech Connect (OSTI)

    Sze, D.K.; Hua, T.Q. [Argonne National Lab., IL (United States); Abdou, M.A. [Univ. of California, Los Angeles, CA (United States); Dagher, M.A.; Waganer, L.M.

    1997-04-01T23:59:59.000Z

    The purpose of the ITER Blanket Testing Module is to test the operating and performance of candidate blanket concepts under a real fusion environment. To assure fuel self-sufficiency the tritium breeding, recovery and processing have to be demonstrated. The tritium produced in the blanket has to be processed to a purity which can be used for refueling. All these functions need to be accomplished so that the tritium system can be scaled to a commercial fusion power plant from a safety and reliability point of view. This paper summarizes the tritium processing steps, the size of the equipment, power requirements, space requirements, etc. for a self-cooled lithium blanket. This information is needed for the design and layout of the test blanket ancillary system and to assure that the ITER guidelines for remote handling of ancillary equipment can be met.

  18. PPPL3188, Preprint: May 1996, UC 420 Measurements of tritium recycling and isotope exchange in TFTR.

    E-Print Network [OSTI]

    is reported in reference 6 and an account of tritium retention, lithium conditioning and advanced tokamak cooled inconel­718 backing plates. The limiter #12; experiences erosion, codeposition of hydr

  19. PPPL-3188, Preprint: May 1996, UC-420 Measurements of tritium recycling and isotope exchange in TFTR.

    E-Print Network [OSTI]

    is reported in reference 6 and an account of tritium retention, lithium conditioning and advanced tokamak cooled inconel-718 backing plates. The limiter #12;experiences erosion, codeposition of hydrogen

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

    E-Print Network [OSTI]

    Rosenberg, Michael Jonathan

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

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

    E-Print Network [OSTI]

    Palmrose, Donald Edwin

    1986-01-01T23:59:59.000Z

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

  2. from Isotope Production Facility

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

    Cancer-fighting treatment gets boost from Isotope Production Facility April 13, 2012 Isotope Production Facility produces cancer-fighting actinium 2:32 Isotope cancer treatment...

  3. Fuel Fabrication Facility

    National Nuclear Security Administration (NNSA)

    Construction of the Mixed Oxide Fuel Fabrication Facility Construction of the Mixed Oxide Fuel Fabrication Facility November 2005 May 2007 June 2008 May 2012...

  4. The requirements for processing tritium recovered from liquid lithium blankets: The blanket interface

    SciTech Connect (OSTI)

    Clemmer, R.G.; Finn, P.A.; Greenwood, L.R.; Grimm, T.L.; Sze, D.K.; Bartlit, J.R.; Anderson, J.L.; Yoshida, H.; Naruse

    1988-03-01T23:59:59.000Z

    We have initiated a study to define a blanket processing mockup for Tritium Systems Test Assembly. Initial evaluation of the requirements of the blanket processing system have been started. The first step of the work is to define the condition of the gaseous tritium stream from the blanket tritium recovery system. This report summarizes this part of the work for one particular blanket concept, i.e., a self-cooled lithium blanket. The total gas throughput, the hydrogen to tritium ratio, the corrosive chemicals, and the radionuclides are defined. The key discoveries are: the throughput of the blanket gas stream (including the helium carrier gas) is about two orders of magnitude higher than the plasma exhaust stream;the protium to tritium ratio is about 1, the deuterium to tritium ratio is about 0.003;the corrosion chemicals are dominated by halides;the radionuclides are dominated by C-14, P-32, and S-35;their is high level of nitrogen contamination in the blanket stream. 77 refs., 6 figs., 13 tabs.

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

    SciTech Connect (OSTI)

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

    2013-11-01T23:59:59.000Z

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

  6. Guide to research facilities

    SciTech Connect (OSTI)

    Not Available

    1993-06-01T23:59:59.000Z

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

  7. MANAGING BERYLLIUM IN NUCLEAR FACILITY APPLICATIONS

    SciTech Connect (OSTI)

    R. Rohe; T. N. Tranter

    2011-12-01T23:59:59.000Z

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

  8. Future Fixed Target Facilities

    SciTech Connect (OSTI)

    Melnitchouk, Wolodymyr

    2009-01-01T23:59:59.000Z

    We review plans for future fixed target lepton- and hadron-scattering facilities, including the 12 GeV upgraded CEBAF accelerator at Jefferson Lab, neutrino beam facilities at Fermilab, and the antiproton PANDA facility at FAIR. We also briefly review recent theoretical developments which will aid in the interpretation of the data expected from these facilities.

  9. Overview of the preliminary safety analysis of the National Ignition Facility

    SciTech Connect (OSTI)

    Brereton, S.; McLouth, L.; Odell, B. [Lawrence Livermore National Lab., CA (United States)] [and others] [Lawrence Livermore National Lab., CA (United States); and others

    1997-06-01T23:59:59.000Z

    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 individual laser beams onto a tiny deuterium-tritium target located at the center of a spherical target chamber. The NIF has been classified as a low hazard, radiological facility on the basis of a preliminary hazards analysis and according to the DOE methodology for facility classification. This requires that a safety analysis report be prepared under DOE Order 5481.1B, Safety Analysis and Review System. A Preliminary Safety Analysis Report (PSAR) has been approved, which documents and evaluates the safety issues associated with the construction, operation, and decommissioning of the NIF. 10 refs., 6 figs., 4 tabs.

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

    SciTech Connect (OSTI)

    Shaver, Mark W.; Lanning, Donald D.

    2010-02-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Not Available

    1988-12-01T23:59:59.000Z

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

  12. Phase 1 Final Report for In-Situ Tritium Beta Detector

    SciTech Connect (OSTI)

    Berthold, J.W.; Jeffers, L.A.

    1998-04-15T23:59:59.000Z

    The objectives of this three-phase project were to design, develop, and demonstrate a monitoring system capable of detecting and quantifying tritium in situ in ground and surface waters, and in water from effluent lines prior to discharge into public waterways. The tritium detection system design is based on measurement of the low energy beta radiation from the radioactive decay of tritium using a special form of scintillating optical fiber directly in contact with the water to be measured. The system consists of the immersible sensor module containing the optical fiber, and an electronics package, connected by an umbilical cable. The system can be permanently installed for routine water monitoring in wells or process or effluent lines, or can be moved from one location to another for survey use. The electronics will read out tritium activity directly in units of pico Curies per liter, with straightforward calibration. In Phase 1 of the project, we characterized the sensitivity of fluor-doped plastic optical fiber to tritium beta radiation. In addition, we characterized the performance of photomultiplier tubes needed for the system. In parallel with this work, we defined the functional requirements, target specifications, and system configuration for an in situ tritium beta detector that would use the fluor-doped fibers as primary sensors of tritium concentration in water. The major conclusions from the characterization work are: A polystyrene optical fiber with fluor dopant concentration of 2% gave best performance. This fiber had the highest dopant concentration of any fibers tested. Stability may be a problem. The fibers exposed to a 22-day soak in 120 F water experienced a 10x reduction in sensitivity. It is not known whether this was due to the build up of a deposit (a potentially reversible effect) or an irreversible process such as leaching of the scintillating dye.

  13. Tritium control and activation in the Pulse*Star reactor

    SciTech Connect (OSTI)

    Blink, J.A.; Hoffman, N.J.

    1983-12-01T23:59:59.000Z

    Pulse*Star is an inertial fusion reactor that uses LiPb coolant in a pool type geometry. LiPb does not release great quantities of chemical energy in a fire, and the pool geometry reduces the difficulty of safely transporting the extremely dense fluid. The compact geometry and good neutronics qualities of LiPb lead to a thermal-to-fusion energy ratio of 1.26, a tritium breeding ratio of 1.22, and a net electric power density 29 times higher than in a fission reactor containment building. The afterheat of the coolant and steel is low enough that emergency cooling systems will be either simple or not required. The gamma dose rate of the bell jar or screen is high enough to require remote maintenance of these components. The steam generators and pumps are on the borderline between limited hands-on and remote maintenance. With additional design attention, limited hands-on maintenance could be feasible for these components. The biological hazard potential indicates that only 10/sup -7/ to 10/sup -6/ of the reactor central region can be vaporized and released; these are values typical of other fusion reactor designs.

  14. Non-conventional passive sensors for monitoring tritium on surfaces

    SciTech Connect (OSTI)

    Gammage, R.B.; Brock, J.L.; Meyer, K.E. [Oak Ridge National Lab., TN (United States). Health Sciences Research Div.

    1995-06-01T23:59:59.000Z

    The authors describe development of small passive, solid-state detectors for in-situ measurements of tritium, or other weak beta-emitting radionuclides, on surfaces. One form of detector operates on the principle of thermally stimulated exoelectron emission (TSEE), the other by discharge of an electret ion chamber (EIC). There are currently two specific types of commercially available detector systems that lend themselves to making surface measurements. One is the thin-film BeO on a graphite disc, and the other is the Teflon EIC. Two other types of TSEE dosimeters (ceramic BeO and carbon doped alumina) are described but lack either a suitable commercially available reader or standardized methods of fabrication. The small size of these detectors allows deployment in locations difficult to access with conventional windowless gas-flow proportional counters. Preliminary testing shows that quantitative measurements are realized with exposure times of 1--10 hours for the TSEE dosimeters (at the DOE release guideline of 5,000 dpm/100 cm{sup 2} for fixed beta contamination). The EIC detectors exhibit an MDA of 26,000 dpm/100 cm{sup 2} for a 24 hour exposure. Both types of integrating device are inexpensive and reusable. Measurements can, therefore, be made that are faster, cheaper, safer, and better than those possible with baseline monitoring technology.

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

    SciTech Connect (OSTI)

    Hysong, R. J.

    1998-12-21T23:59:59.000Z

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

  16. Licensing for tritium production in a commercial light water reactor: A utility view

    SciTech Connect (OSTI)

    Chardos, J.S.; Sorensen, G.C.; Erickson, L.W.

    2000-07-01T23:59:59.000Z

    In a December 1995 Record of Decision for the Final Programmatic Environmental Impact Statement for Tritium Supply and Recycling, the US Department of Energy (DOE) decided to pursue a dual-track approach to determine the preferred option for future production of tritium for the nuclear weapons stockpile. The two options to be pursued were (a) the Accelerator Production of Tritium and (b) the use of commercial light water reactors (CLWRs). DOE committed to select one of these two options as the primary means of tritium production by the end of 1998. The other option would continue to be pursued as a backup to the primary option. The Tennessee Valley Authority (TVA) became involved in the tritium program in early 1996, in response to an inquiry from Pacific Northwest National Laboratory (PNNL) for an expression of interest by utilities operating nuclear power plants (NPPs). In June 1996, TVA was one of two utilities to respond to a request for proposals to irradiate lead test assemblies (LTAs) containing tritium-producing burnable absorber rods (TPBARs). TVA proposed that the LTAs be placed in Watts Bar NPP Unit 1 (WBN). TVA participated with DOE (the Defense Programs Office of CLWR Tritium Production), PNNL, and Westinghouse Electric Company (Westinghouse) in the design process to ensure that the TPBARs would be compatible with safe operation of WBN. Following US Nuclear Regulatory Commission (NRC) issuance of a Safety Evaluation Report (SER) (NUREG-1607), TVA submitted a license amendment request to the NRC for approval to place four LTAs, containing eight TPBARs each, in WBN during the September 1997 refueling outage. In December 1998, DOE announced the selection of the CLWR program as the primary option for tritium production and identified the TVA WBN and Sequoyah NPP (SQN) Units 1 and 2 (SQN-1 and SQN-2, respectively) reactors as the preferred locations to perform tritium production. TVA will prepare license amendment requests for the three plants (WBN, SQN-1, and SQN-2). While the TPBARs replace discreet burnable absorbers in the reactor cores, there are differences in the reactions that occur in the absorber material (lithium aluminate versus boron). At end of life, the lithium aluminate provides considerably more reactivity holddown than the standard boron-containing burnable absorbers. Therefore, it will be necessary for the TVA plant engineering and fuels staffs, working with the fuel vendors, to define the appropriate core loading (number of fresh fuel assemblies, enrichment, etc.) to maintain safe operating limits under both operating and accident conditions. It is recognized that the irradiation of TPBARs in the TVA reactors will also require additional radiological and chemistry program upgrades.

  17. Evaluation of the Fate and Transport of Tritium Contaminated Groundwater from the 618-11 Burial Ground

    SciTech Connect (OSTI)

    Vermeul, Vince R.; Bergeron, Marcel P.; Dresel, P Evan; Freeman, Eugene J.; Peterson, R E.; Thorne, Paul D.

    2005-08-08T23:59:59.000Z

    Tritium transport simulations were conducted to model the mechanisms associated with dilution, dispersion, and radioactive decay that attenuate the 618-11 tritium plume and limit the risk associated with exposure to the Columbia River and Energy Northwest water supply wells. A comparison of simulated and observed tritium concentrations at two downgradient monitoring wells indicated that the model was a reasonable representation of the tritium concentrations immediately downgradient of the site (699-13-3A) and near the leading edge of the plume (699-13-0A). This good match increased confidence in the conceptual model, its numeric implementation, and ultimately, the validity of predictive simulations of tritium fate and transport. Three release scenarios were investigated to measure the impact of the tritium plume at primary receptor locations under different conditions. The three cases were (1) a pulse release of tritium from the burial ground that was the best fit between observed and simulated tritium concentrations; (2) a continuing, decaying source beneath the burial ground through 2015, the milestone for source removal under the River Corridor Closure Contract; and (3) a pulse release as in the best fit case but at twice the concentration. For the best fit case, the model predicts that the maximum tritium concentration will decline to below the drinking water standard by 2031 For the other two release scenarios, maximum tritium concentrations declined to below the drinking water standard by 2040 and 2037, respectively. Tritium from the 618-11 burial ground is not expected to migrate to the Columbia River or to the Energy Northwest water supply wells at concentrations that would pose a significant risk.

  18. Evaluation of the Fate and Transport of Tritium Contaminated Groundwater from the 618-11 Burial Ground

    SciTech Connect (OSTI)

    Vermeul, Vince R.; Bergeron, Marcel P.; Dresel, P EVAN.; Freeman, Eugene J.; Peterson, R E.; Thorne, Paul D.

    2005-10-12T23:59:59.000Z

    Tritium transport simulations were conducted to model the mechanisms associated with dilution, dispersion, and radioactive decay that attenuate the 618-11 Burial Ground tritium plume and limit the risk associated with exposure to the Columbia River and Energy Northwest water supply wells. A comparison of simulated and observed tritium concentrations at two downgradient monitoring wells indicated that the model was a reasonable representation of the tritium concentrations immediately downgradient of the site (699-13-3A) and near the leading edge of the plume (699-13-0A). This good match increased confidence in the conceptual model, its numeric implementation, and ultimately the validity of predictive simulations of tritium fate and transport. Three release scenarios were investigated to measure the impact of the tritium plume at primary receptor locations under different conditions. The three cases were 1) a pulse release of tritium from the burial ground that was the best fit between observed and simulated tritium concentrations; 2) a continuing, decaying source beneath the burial ground through 2015, the milestone for source removal under the River Corridor Closure Contract; and 3) a pulse release as in the best fit case but at twice the concentration. For the best fit case, the model predicts that the maximum tritium concentration will decline to below the drinking water standard by 2031 For the other two release scenarios, maximum tritium concentrations declined to below the drinking water standard by 2040 and 2037, respectively. Tritium from the 618-11 burial ground is not expected to migrate to the Columbia River or to the Energy Northwest water supply wells at concentrations that would pose a significant risk.

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

    SciTech Connect (OSTI)

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

    2008-07-15T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1995-09-01T23:59:59.000Z

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

  1. CRAD, Facility Safety- Nuclear Facility Safety Basis

    Broader source: Energy.gov [DOE]

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

  2. Texas Facilities Commission's Facility Management Strategic Plan

    E-Print Network [OSTI]

    Ramirez, J. A.

    , Texas, November 17 - 19, 2009 Facility Strategic Plan ?High Performance Building Approach ? Envelope ? Load Reduction ? (Re)Design ? Advanced Tactics ?Building Automation ? Sub-metering ? Controls ?Commissioning ? Assessment ? Continuous ?Facility... International Conference for Enhanced Building Operations, Austin, Texas, November 17 - 19, 2009 Commissioning Assessment ?30 buildings ?CC Opportunities ?O&M Improvements ?Energy/Capital Improvement Opportunities ?Quick Payback Implementation ?Levering DM...

  3. Development of Tritium AMS for Biomedical Sciences Research

    SciTech Connect (OSTI)

    Dingley, K H; Chiarappa-Zucca, M L

    2002-01-01T23:59:59.000Z

    Tritium ({sup 3}H) is a radioisotope that is extensively utilized in biological research. Normally in the biological sciences, {sup 3}H is quantified by liquid scintillation counting. For the most sensitive measurements, liquid scintillation counting requires large samples and counting times of several-hours. In contrast, provisional studies at LLNL's Center for Accelerator Mass Spectrometry have demonstrated that Accelerator Mass Spectrometry (AMS) can be-used to quantify {sup 3}H in milligram-sized biological samples with a 100 1000-fold improvement in detection limits when compared to scintillation counting. This increased sensitivity is expected to have great impact in the biological research community. However, before {sup 3}H AMS can be used routinely and successfully, two areas of concern needed to be addressed: (1) sample preparation methods needed to be refined and standardized, and (2) smaller and simpler AMS instrumentation needed to be developed. To address these concerns, the specific aims of this project were to: (1) characterize small dedicated {sup 3}H AMS spectrometer (2) develop routine and robust biological sample preparation methods, and (3) with the aid of our external collaborations, demonstrate the application of {sup 3}H AMS in the biomedical sciences. Towards these goals, the {sup 3}H AMS instrument was installed and optimized to enhance performance. The sample preparation methodology was established for standard materials (water and tributyrin) and biological samples. A number of biological and environmental studies which require {sup 3}H AMS were undertaken with university collaborators and our optimized analysis methods were employed to measure samples from these projects.

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

    E-Print Network [OSTI]

    Abdou, Mohamed

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

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

    E-Print Network [OSTI]

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

  6. PPPL-3300, Preprint: May 1998, UC-420,423 Tritium Experience in Large Tokamaks: Application to ITER

    E-Print Network [OSTI]

    experience with the use of tritium fuel in the Tokamak Fusion Test Reactor and the Joint European Torus, together with progress in developing the technical design of the International Thermonuclear Experimental Reactor has expanded the technical knowledge base for tritium issues in fusion. This paper reports

  7. PPPL3300, Preprint: May 1998, UC420,423 Tritium Experience in Large Tokamaks: Application to ITER

    E-Print Network [OSTI]

    experience with the use of tritium fuel in the Tokamak Fusion Test Reactor and the Joint European Torus, together with progress in developing the technical design of the International Thermonuclear Experimental Reactor has expanded the technical knowledge base for tritium issues in fusion. This paper reports

  8. Extraction efficiency and quantification of mutagenic chemicals in soils

    E-Print Network [OSTI]

    Maggard, Lea Ann

    1986-01-01T23:59:59.000Z

    : Dr. K. W. Brown Lack of established extraction procedures for quantification of mutagenic compounds in soil is a major technical limitation to monitoring and assessing the performance of a hazardous waste land treatment facility. In this study... for extracting organic mutagens from the waste or soil/waste mixture. The use of combined biological and chemical testing protocol provided the most practical means of determining extraction efficiency. The bioassay detected additive, synergistic...

  9. Technology Transitions Facilities Database

    Broader source: Energy.gov [DOE]

    The types of R&D facilities at the DOE Laboratories available to the public typically fall into three broad classes depending on the mode of access: Designated User Facilities, Shared R&D...

  10. Utilization of kinetic isotope effects for the concentration of tritium. 1997 annual progress report

    SciTech Connect (OSTI)

    Brown, G.M.; Meyer, T.J.

    1997-09-01T23:59:59.000Z

    '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 tritiated water (HTO)] is in a rapidly established equilibrium with water, it moves with groundwater and separation from water cannot be achieved by the usual pump-and-treat methods using sorbents. The general methodology developed in this work will be applicable to a number of US Department of Energy waste streams, and as a consequence of the process, tritium could be incorporated in an organic polymer, a form that will prevent its ready transport in groundwater.'

  11. KATRIN: an experiment to determine the neutrino mass from the beta decay of tritium

    E-Print Network [OSTI]

    R. G. Hamish Robertson

    2013-07-21T23:59:59.000Z

    KATRIN is a very large scale tritium-beta-decay experiment to determine the mass of the neutrino. It is presently under construction at the Karlsruhe Institute of Technology north campus, and makes use of the Karlsruhe Tritium Laboratory built as a prototype for the ITER project. The combination of a large retarding-potential electrostatic-magnetic spectrometer and an intense gaseous molecular tritium source makes possible a sensitivity to neutrino mass of 0.2 eV, about an order of magnitude below present laboratory limits. The measurement is kinematic and independent of whether the neutrino is Dirac or Majorana. The status of the project is summarized briefly in this report.

  12. Role of Sterile Neutrino Warm Dark Matter in Rhenium and Tritium Beta Decays

    E-Print Network [OSTI]

    H. J. de Vega; O. Moreno; E. Moya de Guerra; M. Ramon Medrano; N. Sanchez

    2012-09-24T23:59:59.000Z

    Sterile neutrinos with mass in the range of one to a few keV are important as extensions of the Standard Model of particle physics and are serious dark matter (DM) candidates. This DM mass scale (warm DM) is in agreement with both cosmological and galactic observations. We study the role of a keV sterile neutrino through its mixing with a light active neutrino in Rhenium 187 and Tritium beta decays. We pinpoint the energy spectrum of the beta particle, 0 Tritium beta spectra and estimate the size of this perturbation by means of the dimensionless ratio R of the sterile neutrino to the active neutrino contributions. We comment on the possibility of searching for sterile neutrino signatures in two experiments which are currently running at present, MARE and KATRIN, focused on the Rhenium 187 and Tritium beta decays respectively.

  13. KATRIN: an experiment to determine the neutrino mass from the beta decay of tritium

    E-Print Network [OSTI]

    ,

    2013-01-01T23:59:59.000Z

    KATRIN is a very large scale tritium-beta-decay experiment to determine the mass of the neutrino. It is presently under construction at the Karlsruhe Institute of Technology north campus, and makes use of the Karlsruhe Tritium Laboratory built as a prototype for the ITER project. The combination of a large retarding-potential electrostatic-magnetic spectrometer and an intense gaseous molecular tritium source makes possible a sensitivity to neutrino mass of 0.2 eV, about an order of magnitude below present laboratory limits. The measurement is kinematic and independent of whether the neutrino is Dirac or Majorana. The status of the project is summarized briefly in this report.

  14. Tests of Tritium Pellet Injector TPI-1 Under Closed Cycle Mode

    SciTech Connect (OSTI)

    Vedeneev, A.I. [Russian Federal Nuclear Center-All-Russia Scientific Institute of Experimental Physics (Russian Federation); Abramov, I.A. [Russian Federal Nuclear Center-All-Russia Scientific Institute of Experimental Physics (Russian Federation); Lebedev, S.E. [Russian Federal Nuclear Center-All-Russia Scientific Institute of Experimental Physics (Russian Federation); Kazakovsky, N.T. [Russian Federal Nuclear Center-All-Russia Scientific Institute of Experimental Physics (Russian Federation); Pimanikhin, S.A. [Russian Federal Nuclear Center-All-Russia Scientific Institute of Experimental Physics (Russian Federation); Saksagansky, G.L. [Efremov Institute (Russian Federation); Sten'gach, A.V. [Russian Federal Nuclear Center-All-Russia Scientific Institute of Experimental Physics (Russian Federation); Shirnin, P.V. [Russian Federal Nuclear Center-All-Russia Scientific Institute of Experimental Physics (Russian Federation); Viniar, I.V

    2005-07-15T23:59:59.000Z

    First experimental results of a tritium pellet injector steady-state operation is presented. The tritium injector TPI-1 was developed at the PELIN Laboratory and put in operation in Russian Federal Nuclear Center. It is a part of an experimental closed circuit for simulation of ITER fuel cycle. Results of several continuous extrusions of solid rod made of various hydrogen isotopes are presented. Transverse dimensions of an extruded ice rod with rectangular cross-section were {approx} 3 x 4.mm. The greatest extrusion velocity came to 15 mm/s for hydrogen and 9 mm/s for D-T mixture; tritium content in fuel mixture did not exceed 11%; pellet velocity ran up to 500 m/s at repetitive mode. An optimal mode of D-T ice extrusion was determined.

  15. Apparatus for hydrocarbon extraction

    DOE Patents [OSTI]

    Bohnert, George W.; Verhulst, Galen G.

    2013-03-19T23:59:59.000Z

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

  16. PRODUCTION DE TRITIUM DANS LE THORIUM PAR DES PROTONS DE 135 MeV Par M. LEFORT, G. SIMONOFF et X. TARRAGO

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    959 PRODUCTION DE TRITIUM DANS LE THORIUM PAR DES PROTONS DE 135 MeV Par M. LEFORT, G. SIMONOFF et thorium par des protons de 135 MeV accélérés au synchro-cyclotron d'Orsay. Le tritium était extrait des measured the cross-section of tritium production by bombardement of thorium by 135 MeV protons in the Orsay

  17. Physics of Aquatic Systems II, 7. Tritium and Helium-3 Universitt HeidelbergInstitut fr Umweltphysik Physics of Aquatic Systems II

    E-Print Network [OSTI]

    Aeschbach-Hertig, Werner

    Physics of Aquatic Systems II, 7. Tritium and Helium-3 Universität HeidelbergInstitut für Umweltphysik 1 Physics of Aquatic Systems II ­ 7. Tritium and Helium-3 Werner Aeschbach-Hertig Institute of Environmental Physics University of Heidelberg Physics of Aquatic Systems II, 7. Tritium and Helium-3

  18. Proceedings of SOFE97 17th IEEE/NPSS Symposium on Fusion Engineering, Oct.6-9th, 1997. Tritium Removal by CO2 Laser Heating*.

    E-Print Network [OSTI]

    Proceedings of SOFE97 17th IEEE/NPSS Symposium on Fusion Engineering, Oct.6-9th, 1997. Tritium, NM 87185 Abstract -- Efficient techniques for rapid tritium removal will be necessary for ITER by a scanning CO2 or Nd:Yag laser that would release tritium without the severe engineering difficulties of bulk

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

    SciTech Connect (OSTI)

    Kapyshev, Victor K.; Chernetsov, Mikhail Yu.; Zhevotov, Sergej I.; Kersnovskij, Alexandr Yu.; Kolbasov, Boris N.; Kovalenko, Victor G.; Paltusov, Nikolaj P.; Sernyaev, Georgeij A.; Sterebkov, Juri S.; Zyryanov, Alexej P. [A.A. Bochvar Institute of Inorganic Materials (Russian Federation)

    2005-07-15T23:59:59.000Z

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

  20. Proceedings of the 4th International Workshop on Tritium Effects in Plasma Facing Components

    SciTech Connect (OSTI)

    R. A. Causey

    1999-02-01T23:59:59.000Z

    The 4th International Workshop on Tritium Effects in Plasma Facing Components was held in Santa Fe, New Mexico on May 14-15, 1998. This workshop occurs every two years, and has previously been held in Livermore/California, Nagoya/Japan, and the JRC-Ispra Site in Italy. The purpose of the workshop is to gather researchers involved in the topic of tritium migration, retention, and recycling in materials used to line magnetic fusion reactor walls and provide a forum for presentation and discussions in this area. This document provides an overall summary of the workshop, the workshop agenda, a summary of the presentations, and a list of attendees.

  1. Results of a baseflow tritium survey of surface water in Georgia across from the Savannah River Site

    SciTech Connect (OSTI)

    Nichols, R.L.

    1993-03-03T23:59:59.000Z

    In October 1991 the Georgia Department of Natural Resources (GDNR) issued a press release notifying the public that tritium had been measured in elevated levels (1,200 - 1,500 pCi/1) in water samples collected from drinking water wells in Georgia across from the Savannah River Site in Aiken Co. South Carolina. None of the elevated results were above the Primary Drinking Water Standard for tritium of 20,000 pCi/l. The GDNR initiated 2 surveys to determine the source and extent of elevated tritium: (1) baseflow survey of surface water quality, and (2) well evaluation program. Results from the 2 surveys indicate that the tritium measured in groundwater wells in Georgia is not the result of a groundwater flow from South Carolina under the Savannah River and into Georgia. Atmospheric transport and consequent rainout and infiltration has resulted in an increase of tritium in the water-table aquifer in the vicinity. Water samples collected from drinking water wells believed to have been installed in the aquifer beneath the water-table aquifer were actually from the shallower water-table aquifer. Water samples collected from the wells contain the amount of tritium expected for the water-table aquifer in the sample area. The measured tritium levels in the well samples and baseflow samples do not exceed Primary Drinking Water Standards. Tritium levels in the water-table in Georgia will decline as the atmospheric releases from SRS decline, tritium undergoes natural decay, and infiltration water with less tritium flushes through the subsurface.

  2. Small Power Production Facilities (Montana)

    Broader source: Energy.gov [DOE]

    For the purpose of these regulations, a small power production facility is defined as a facility that:...

  3. Stockpile Stewardship and the National Ignition Facility

    SciTech Connect (OSTI)

    Moses, E

    2012-01-04T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Ross, P.

    2012-08-29T23:59:59.000Z

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

  5. Waste Management facilities fault tree databank 1995 status report

    SciTech Connect (OSTI)

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

    1995-08-16T23:59:59.000Z

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

  6. Summary of the IEA Workshop on Alpha Physics and Tritium Issues in Large Tokamaks

    SciTech Connect (OSTI)

    Cheng, C.Z.; Stratton, B.; Zweben, S.J. [Princeton Univ., NJ (United States). Plasma Physics Lab.; Pitcher, C.S. [Toronto Univ., Downsview, ON (Canada)

    1993-11-01T23:59:59.000Z

    A brief summary is presented of the talks given during this meeting, which was held at PPPL and sponsored by the IEA (International Energy Agency) as part of the Large Tokamak collaboration. These talks are summarized into four sessions: tritium issues in large tokamaks, alpha particle simulation experiments, alpha particle theory, and alpha particle diagnostics.

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

    E-Print Network [OSTI]

    Abdou, Mohamed

    , Westinghouse Hanford Company C. E. Johnson, Argonne National Laboratory M. Abdou, University of California. DISCLAIMER This report was prepared as an account of work sponsored by an agency of the United States which can potentially satisfy the challenging engineering requirements of tritium producing blankets

  8. Neutron profile measurements for trace tritium experiments M. J. Loughlin, N. Watkins,a)

    E-Print Network [OSTI]

    Basse, Nils Plesner

    system, its operation and assesses the difficulties due to scattered neutrons. The profiles can be usedNeutron profile measurements for trace tritium experiments M. J. Loughlin, N. Watkins,a) J. M, United Kingdom Presented on 11 June 1998 The JET neutron profile monitor was used to study the transport

  9. A vacuum disengager for tritium removal from HYLIFE-II Reactor Flibe

    SciTech Connect (OSTI)

    Dolan, T.J.; Longhurst, G.R. (EG and G Idaho, Inc., Idaho Falls, ID (United States)); Garcia-Otero, E. (Missouri Univ., Columbia, MO (United States). Dept. of Nuclear Engineering)

    1992-01-01T23:59:59.000Z

    We have designed a vacuum disengager system to remove tritium from the Flibe (Li{sub 2}BeF{sub 4}) molten salt coolant of the HYLIFE-II fusion reactor. There is a two-stage vacuum disengager in each of three intermediate heat exchanger (IHX) loops. Each stage consists of a vacuum chamber 4 m in diameter and 7 m tall. As 0.2 mm diameter molten salt droplets fall vertically downward into the vacuum, most of the tritium diffuses out of the droplets and is pumped away. A fraction {Phi} {approximately}10{sup {minus}5} of the 8.6 MCi/day tritium source (from breeding in the Flibe and from unburned fuel) remains in the Flibe as it leaves the vacuum disengagers, and about 21% of that permeates into the intermediate coolant loop, so about 20 Ci/day leak into the steam system. With Flibe primary coolant and a vacuum disengager, it appears that an intermediate coolant loop is not needed to prevent tritium from leaking into the steam system. An experiment is needed to demonstrate Flibe vacuum disengager operation.

  10. A vacuum disengager for tritium removal from HYLIFE-II Reactor Flibe

    SciTech Connect (OSTI)

    Dolan, T.J.; Longhurst, G.R. [EG and G Idaho, Inc., Idaho Falls, ID (United States); Garcia-Otero, E. [Missouri Univ., Columbia, MO (United States). Dept. of Nuclear Engineering

    1992-09-01T23:59:59.000Z

    We have designed a vacuum disengager system to remove tritium from the Flibe (Li{sub 2}BeF{sub 4}) molten salt coolant of the HYLIFE-II fusion reactor. There is a two-stage vacuum disengager in each of three intermediate heat exchanger (IHX) loops. Each stage consists of a vacuum chamber 4 m in diameter and 7 m tall. As 0.2 mm diameter molten salt droplets fall vertically downward into the vacuum, most of the tritium diffuses out of the droplets and is pumped away. A fraction {Phi} {approximately}10{sup {minus}5} of the 8.6 MCi/day tritium source (from breeding in the Flibe and from unburned fuel) remains in the Flibe as it leaves the vacuum disengagers, and about 21% of that permeates into the intermediate coolant loop, so about 20 Ci/day leak into the steam system. With Flibe primary coolant and a vacuum disengager, it appears that an intermediate coolant loop is not needed to prevent tritium from leaking into the steam system. An experiment is needed to demonstrate Flibe vacuum disengager operation.

  11. TRITIUM UNCERTAINTY ANALYSIS FOR SURFACE WATER SAMPLES AT THE SAVANNAH RIVER SITE

    SciTech Connect (OSTI)

    Atkinson, R.

    2012-07-31T23:59:59.000Z

    Radiochemical analyses of surface water samples, in the framework of Environmental Monitoring, have associated uncertainties for the radioisotopic results reported. These uncertainty analyses pertain to the tritium results from surface water samples collected at five locations on the Savannah River near the U.S. Department of Energy's Savannah River Site (SRS). Uncertainties can result from the field-sampling routine, can be incurred during transport due to the physical properties of the sample, from equipment limitations, and from the measurement instrumentation used. The uncertainty reported by the SRS in their Annual Site Environmental Report currently considers only the counting uncertainty in the measurements, which is the standard reporting protocol for radioanalytical chemistry results. The focus of this work is to provide an overview of all uncertainty components associated with SRS tritium measurements, estimate the total uncertainty according to ISO 17025, and to propose additional experiments to verify some of the estimated uncertainties. The main uncertainty components discovered and investigated in this paper are tritium absorption or desorption in the sample container, HTO/H{sub 2}O isotopic effect during distillation, pipette volume, and tritium standard uncertainty. The goal is to quantify these uncertainties and to establish a combined uncertainty in order to increase the scientific depth of the SRS Annual Site Environmental Report.

  12. UNCLASSIFIED TPBAR RELEASES, INCLUDING TRITIUM TTQP-1-091 Rev 14

    SciTech Connect (OSTI)

    Gruel, Robert L.; Love, Edward F.; Thornhill, Cheryl K.

    2012-07-01T23:59:59.000Z

    This document provides a listing of unclassified tritium release values that should be assumed for unclassified analysis. Much of the information is brought forth from the related documents listed in Section 5.0 to provide a single-source listing of unclassified release values. This information has been updated based on current design analysis and available experimental data.

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

    E-Print Network [OSTI]

    Andrew J. Long; Cecilia Lunardini; Eray Sabancilar

    2014-11-12T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    1 PPPL­3157 ­ Preprint Date: March 1996, UC­421, 423, 426 Investigations of the Tritium Recycling material to be ejected into the plasma. This recycling of plasma fuel, which occurs primarily on the inner influx from the edge. Despite its importance, a full understanding of the factors influencing recycling

  15. PPPL-3157 -Preprint Date: March 1996, UC-421, 423, 426 Investigations of the Tritium Recycling

    E-Print Network [OSTI]

    1 PPPL-3157 - Preprint Date: March 1996, UC-421, 423, 426 Investigations of the Tritium Recycling material to be ejected into the plasma. This recycling of plasma fuel, which occurs primarily on the inner influx from the edge. Despite its importance, a full understanding of the factors influencing recycling

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

    E-Print Network [OSTI]

    Abdou, Mohamed

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

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

    SciTech Connect (OSTI)

    Nickels, J.M.

    1991-08-01T23:59:59.000Z

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

  18. Heat Exchanger Design Options and Tritium Transport Study for the VHTR System

    SciTech Connect (OSTI)

    Chang H. Oh; Eung S. Kim

    2008-09-01T23:59:59.000Z

    This report presents the results of a study conducted to consider heat exchanger options and tritium transport in a very high temperature reactor (VHTR) system for the Next Generation Nuclear Plant Project. The heat exchanger options include types, arrangements, channel patterns in printed circuit heat exchangers (PCHE), coolant flow direction, and pipe configuration in shell-and-tube designs. Study considerations include: three types of heat exchanger designs (PCHE, shell-and-tube, and helical coil); single- and two-stage unit arrangements; counter-current and cross flow configurations; and straight pipes and U-tube designs in shell-and-tube type heat exchangers. Thermal designs and simple stress analyses were performed to estimate the heat exchanger options, and the Finite Element Method was applied for more detailed calculations, especially for PCHE designs. Results of the options study show that the PCHE design has the smallest volume and heat transfer area, resulting in the least tritium permeation and greatest cost savings. It is theoretically the most reliable mechanically, leading to a longer lifetime. The two-stage heat exchanger arrangement appears to be safer and more cost effective. The recommended separation temperature between first and second stages in a serial configuration is 800oC, at which the high temperature unit is about one-half the size of the total heat exchanger core volume. Based on simplified stress analyses, the high temperature unit will need to be replaced two or three times during the plant’s lifetime. Stress analysis results recommend the off-set channel pattern configuration for the PCHE because stress reduction was estimated at up to 50% in this configuration, resulting in a longer lifetime. The tritium transport study resulted in the development of a tritium behavior analysis code using the MATLAB Simulink code. In parallel, the THYTAN code, previously performed by Ohashi and Sherman (2007) on the Peach Bottom data, was revived and verified. The 600 MWt VHTR core input file developed in preparation for the transient tritium analysis of VHTR systems was replaced with the original steady-state inputs for future calculations. A Finite Element Method analysis was performed using COMSOL Multiphysics software to accurately predict tritium permeation through the PCHE type heat exchanger walls. This effort was able to estimate the effective thickness for tritium permeations and develop a correlation for general channel configurations, which found the effective thickness to be much shorter than the average channel distance because of dead spots on the channel side.

  19. Overview of the Defense Programs Research and Technology Development Program for fiscal year 1993. Appendix II research laboratories and facilities

    SciTech Connect (OSTI)

    Not Available

    1993-09-30T23:59:59.000Z

    This document contains summaries of the research facilities that support the Defense Programs Research and Technology Development Program for FY 1993. The nine program elements are aggregated into three program clusters as follows: (1) Advanced materials sciences and technologies; chemistry and materials, explosives, special nuclear materials (SNM), and tritium. (2) Design sciences and advanced computation; physics, conceptual design and assessment, and computation and modeling. (3) Advanced manufacturing technologies and capabilities; system engineering science and technology, and electronics, photonics, sensors, and mechanical components. Section I gives a brief summary of 23 major defense program (DP) research and technology facilities and shows how these major facilities are organized by program elements. Section II gives a more detailed breakdown of the over 200 research and technology facilities being used at the Laboratories to support the Defense Programs mission.

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

    SciTech Connect (OSTI)

    C. Cooper; M. Ye; J. Chapman

    2008-04-01T23:59:59.000Z

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

  1. Independent Oversight Review, Savannah River Field Office Tritium

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels DataDepartment of Energy Your Density Isn't YourTransport(Fact Sheet),EnergyImprovement ofDecemberPlateau Remediation CompanyFacilities

  2. Dosimetric impact evaluation of primary coolant chemistry of the internal tritium breeding cycle of a fusion reactor DEMO

    SciTech Connect (OSTI)

    Velarde, M. [Instituto de Fusion Nuclear (DENIM), ETSII, Universidad Politecnica Madrid UPM, J. Gutierrez Abascal 2, Madrid 28006 (Spain); Sedano, L. A. [Asociacion Euratom-Ciematpara Fusion, Av. Complutense 22, 28040 Madrid (Spain); Perlado, J. M. [Instituto de Fusion Nuclear (DENIM), ETSII, Universidad Politecnica Madrid UPM, J. Gutierrez Abascal 2, Madrid 28006 (Spain)

    2008-07-15T23:59:59.000Z

    Tritium will be responsible for a large fraction of the environmental impact of the first generation of DT fusion reactors. Today, the efforts of conceptual development of the tritium cycle for DEMO are mainly centred in the so called Inner Breeding Tritium Cycle, conceived as guarantee of reactor fuel self-sufficiency. The EU Fusion Programme develops for the short term of fusion power technology two breeding blanket conceptual designs both helium cooled. One uses Li-ceramic material (HCPB, Helium-Cooled Pebble Bed) and the other a liquid metal eutectic alloy (Pb15.7Li) (HCLL, Helium-Cooled Lithium Lead). Both are Li-6 enriched materials. At a proper scale designs will be tested as Test Blanket Modules in ITER. The tritium cycles linked to both blanket concepts are similar, with some different characteristics. The tritium is recovered from the He purge gas in the case of HCPB, and directly from the breeding alloy through a carrier gas in HCLL. For a 3 GWth self-sufficient fusion reactor the tritium breeding need is few hundred grams of tritium per day. Safety and environmental impact are today the top priority design criteria. Dose impact limits should determine the key margins and parameters in its conception. Today, transfer from the cycle to the environment is conservatively assumed to be operating in a 1-enclosure scheme through the tritium plant power conversion system (intermediate heat exchangers and helium blowers). Tritium loss is caused by HT and T{sub 2} permeation and simultaneous primary coolant leakage through steam generators. Primary coolant chemistry appears to be the most natural way to control tritium permeation from the breeder into primary coolant and from primary coolant through SG by H{sub 2} tritium flux isotopic swamping or steel (EUROFER/INCOLOY) oxidation. A primary coolant chemistry optimization is proposed. Dynamic flow process diagrams of tritium fluxes are developed ad-hoc and coupled with tritiated effluents dose impact evaluations. Dose assessments are obtained from the use of appropriate numeric tools (NORMTRI). (authors)

  3. METHOD FOR IN SITU VISUALIZATION OF TRITIUM DIFFUSED IN STAINLESS STEEL USING A DIGITAL AUTORADIOGRAPHIC IMAGING SYSTEM

    SciTech Connect (OSTI)

    Gibbs, K; Carol Kestin, C

    2006-09-20T23:59:59.000Z

    At the end of their service lives, various stainless steel components of nuclear weapons that have been exposed to tritium gas are evaluated to determine the extent of the tritium permeation. This information is then used to assess the decrement to performance caused by hydrogen (tritium) embrittlement. This evaluation is currently performed using a photo-emulsion based method and requires 24 hours or longer to complete. A system based on digital imaging technology has recently been designed and built at the Savannah River National Laboratory that performs this evaluation in 10 minutes or less on typical samples.

  4. ARM Mobile Facilities

    ScienceCinema (OSTI)

    Orr, Brad; Coulter, Rich

    2014-09-15T23:59:59.000Z

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

  5. DOE Designated Facilities

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

    Reactor** Lawrence Berkeley National Laboratory Joint Genome Institute - Production Genomics Facility (PGF)** (joint with LLNL, LANL, ORNL and PNNL) Advanced Light Source (ALS)...

  6. Accelerator Test Facility

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

    Test Facility Vitaly Yakimenko October 6-7, 2010 ATF User meeting DOE HE, S. Vigdor, ALD - (Contact) T. Ludlam Chair, Physics Department V. Yakimenko Director ATF, Accelerator...

  7. ACCELERATOR TEST FACILITY

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

    LABORATORY PHYSICS DEPARTMENT Effective: 04012004 Page 1 of 2 Subject: Accelerator Test Facility - Linear Accelerator General Systems Guide Prepared by: Michael Zarcone...

  8. Carbon Fiber Technology Facility

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

    The Carbon Fiber Technology Facility is relevant in proving the scale- up of low-cost carbon fiber precursor materials and advanced manufacturing technologies * Significant...

  9. Science and Technology Facility

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

    IBRF Project Lessons Learned Report Integrated Biorefinery Research Facility Lessons Learned - Stage I Acquisition through Stage II Construction Completion August 2011 This...

  10. Programs & User Facilities

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

    Research Facility Climate, Ocean, and Sea Ice Modeling (COSIM) Terrestrial Ecosystem and Climate Dynamics Fusion Energy Sciences Magnetic Fusion Experiments Plasma Surface...

  11. Facilities | Argonne National Laboratory

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

    Some of the nation's most powerful and sophisticated facilities for energy research Argonne National Laboratory is home to some of the nation's most powerful and sophisticated...

  12. Existing Facilities Program

    Broader source: Energy.gov [DOE]

    The NYSERDA Existing Facilities program merges the former Peak Load Reduction and Enhanced Commercial and Industrial Performance programs. The new program offers a broad array of different...

  13. Facility Survey & Transfer

    Broader source: Energy.gov [DOE]

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

  14. Sandia National Laboratories: Facilities

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

    InstituteSandia Photovoltaic Systems Symposium On April 15, 2014, in Concentrating Solar Power, Distribution Grid Integration, Energy, Facilities, Grid Integration, News,...

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

    SciTech Connect (OSTI)

    Sullivan, N.

    1995-05-02T23:59:59.000Z

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

  16. Catalytic extraction processing of contaminated scrap metal

    SciTech Connect (OSTI)

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

    1995-12-01T23:59:59.000Z

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

  17. MINERAL FACILITIES MAPPING PROJECT

    E-Print Network [OSTI]

    Gilbes, Fernando

    MINERAL FACILITIES MAPPING PROJECT Yadira Soto-Viruet Supervisor: David Menzie, Yolanda Fong-Sam Minerals Information Team (MIT) USGS Summer Internship 2009 U.S. Department of the Interior U.S. Geological Minerals Information Team (MIT): Annually reports on the minerals facilities of more than 180 countries

  18. A Materials Facilities Initiative -

    E-Print Network [OSTI]

    A Materials Facilities Initiative - FMITS & MPEX D.L. Hillis and ORNL Team Fusion & Materials for Nuclear Systems Division July 10, 2014 #12;2 Materials Facilities Initiative JET ITER FNSF Fusion Reactor Challenges for materials: fluxes and fluence, temperatures 50 x divertor ion fluxes up to 100 x neutron

  19. Geophysical InversionFacility

    E-Print Network [OSTI]

    Oldenburg, Douglas W.

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

  20. Argonne Leadership Computing Facility

    E-Print Network [OSTI]

    Kemner, Ken

    Argonne Leadership Computing Facility Argonne Leadership Computing Facility 2010 ANNUAL REPORT S C I E N C E P O W E R E D B Y S U P E R C O M P U T I N G ANL-11/15 The Argonne Leadership Computing States Government nor any agency thereof, nor UChicago Argonne, LLC, nor any of their employees

  1. Emergency Facilities and Equipment

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

    1997-08-21T23:59:59.000Z

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

  2. Nanotechnology User Facility for

    E-Print Network [OSTI]

    A National Nanotechnology User Facility for Industry Academia Government #12;The National Institute of Commerce's nanotechnology user facility. The CNST enables innovation by providing rapid access to the tools new measurement and fabrication methods in response to national nanotechnology needs. www

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

    SciTech Connect (OSTI)

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

    2012-02-19T23:59:59.000Z

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

  4. Evaluation of Elevated Tritium Levels in Groundwater Downgradient from the 618-11 Burial Ground Phase I Investigation

    SciTech Connect (OSTI)

    Dresel, P Evan; Williams, Bruce A.; Evans, John C.; Smith, Ronald M.; Thompson, Christopher J.; Hulstrom, Larry C.

    2000-06-06T23:59:59.000Z

    This report describes the results of the preliminary investigation of elevated tritium in groundwater discovered near the 618-11 burial ground, located in the eastern part of the Hanford Site.

  5. Evaluation of Elevated Tritium Levels in Groundwater Downgradient from the 618-11 Burial Ground Phase I Investigations

    SciTech Connect (OSTI)

    Dresel, P Evan; Evans, John C; Hulstrom, Larry C; Smith, Ronald M; Thompson, Christopher J; Williams, Bruce A

    2000-06-06T23:59:59.000Z

    This report describes the results of the preliminary investigation of elevated tritium in groundwater discovered near the 618-11 burial ground, located in the eastern part of the Hanford Site.

  6. PPPL-3279, Preprint: January 1998, UC-420, 426 Toroidal Alfvn Eigenmodes in TFTR Deuterium-Tritium Plasmas

    E-Print Network [OSTI]

    -Tritium Plasmas R. Nazikian, G.Y. Fu, Z. Chang, S.H. Batha1 , H. Berk2 , R.V. Budny, Y. Chen, C.Z. Cheng, D

  7. Demonstration of the Highest Deuterium-Tritium Areal Density Using Multiple-Picket Cryogenic Designs on OMEGA

    E-Print Network [OSTI]

    Goncharov, V. N.

    The performance of triple-picket deuterium-tritium cryogenic target designs on the OMEGA Laser System [T.R. Boehly et al., Opt. Commun. 133, 495 (1997)] is reported. These designs facilitate control of shock heating in ...

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

    E-Print Network [OSTI]

    Rodriguez, Judy N

    2013-01-01T23:59:59.000Z

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

  9. Monitoring of tritium purity during long-term circulation in the KATRIN test experiment LOOPINO using laser Raman spectroscopy

    E-Print Network [OSTI]

    Fischer, Sebastian; Schlösser, Magnus; Bornschein, Beate; Drexlin, Guido; Priester, Florian; Lewis, Richard J; Telle, Helmut H

    2012-01-01T23:59:59.000Z

    The gas circulation loop LOOPINO has been set up and commissioned at Tritium Laboratory Karlsruhe (TLK) to perform Raman measurements of circulating tritium mixtures under conditions similar to the inner loop system of the neutrino-mass experiment KATRIN, which is currently under construction. A custom-made interface is used to connect the tritium containing measurement cell, located inside a glove box, with the Raman setup standing on the outside. A tritium sample (purity > 95%, 20 kPa total pressure) was circulated in LOOPINO for more than three weeks with a total throughput of 770 g of tritium. Compositional changes in the sample and the formation of tritiated and deuterated methanes CT_(4-n)X_n (X=H,D; n=0,1) were observed. Both effects are caused by hydrogen isotope exchange reactions and gas-wall interactions, due to tritium {\\beta} decay. A precision of 0.1% was achieved for the monitoring of the T_2 Q_1-branch, which fulfills the requirements for the KATRIN experiment and demonstrates the feasibility ...

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

    SciTech Connect (OSTI)

    Clark, E; Marie Kane, M

    2008-12-12T23:59:59.000Z

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

  11. DOE/NNSA Facility Management Contracts Facility Owner Contractor

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

    NNSA Facility Management Contracts Facility Owner Contractor Award Date End Date OptionsAward Term Ultimate Potential Expiration Date Contract FY Competed Parent Companies LLC...

  12. Test Facility Daniil Stolyarov, Accelerator Test Facility User...

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

    Development of the Solid-State Laser System for the Accelerator Test Facility Daniil Stolyarov, Accelerator Test Facility User's Meeting April 3, 2009 Outline Motivation for...

  13. Sandia Energy - About the Facility

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

    the Facility About the FacilityTara Camacho-Lopez2015-05-11T19:38:37+00:00 Test-Bed Wind Turbines Allow Facility Flexibility While Providing Reliable Data in Many Regimes SWiFT...

  14. Suppression of Tritium Retention in Remote Areas of ITER by Nonperturbative Reactive Gas Injection

    SciTech Connect (OSTI)

    Tabares, F. L.; Ferreira, J. A.; Ramos, A. [As Euratom-Ciemat, Av Complutense 22, 28040 Madrid (Spain); Rooij, G. van; Westerhout, J.; Al, R.; Rapp, J. [FOM Instituut voor Plasmafysica Rijnhuizen, EURATOM Association, TEC, PO Box 1207, 3430 BE Nieuwegein (Netherlands); Drenik, A.; Mozetic, M. [As Euratom-MHEST, Institut Jozef Stefan, Jamova cesta 39, 1000 Ljubljana (Slovenia)

    2010-10-22T23:59:59.000Z

    A technique based on reactive gas injection in the afterglow region of the divertor plasma is proposed for the suppression of tritium-carbon codeposits in remote areas of ITER when operated with carbon-based divertor targets. Experiments in a divertor simulator plasma device indicate that a 4 nm/min deposition can be suppressed by addition of 1 Pa{center_dot}m{sup 3} s{sup -1} ammonia flow at 10 cm from the plasma. These results bolster the concept of nonperturbative scavenger injection for tritium inventory control in carbon-based fusion plasma devices, thus paving the way for ITER operation in the active phase under a carbon-dominated, plasma facing component background.

  15. Wavelet Approach to Search for Sterile Neutrinos in Tritium $?$-Decay Spectra

    E-Print Network [OSTI]

    S. Mertens; K. Dolde; M. Korzeczek; F. Glueck; S. Groh; R. D. Martin; A. W. P. Poon; M. Steidl

    2015-01-08T23:59:59.000Z

    Sterile neutrinos in the mass range of a few keV are candidates for both cold and warm dark matter. An ad-mixture of a heavy neutrino mass eigenstate to the electron neutrino would result in a minuscule distortion - a 'kink' - in a $\\beta$-decay spectrum. In this paper we show that a wavelet transform is a very powerful shape analysis method to detect this signature. For a tritium source strength, similar to what is expected from the KATRIN experiment, a statistical sensitivity to active-to-sterile neutrino mixing down to $\\sin^2 \\theta= 10^{-6}$ ($90\\%$ CL) can be obtained after 3 years of measurement time. It is demonstrated that the wavelet approach is largely insensitive to systematic effects that result in smooth spectral modifications. To make full use of this analysis technique a high resolution measurement (FWHM of $\\sim100$~eV) of the tritium $\\beta$-decay spectrum is required.

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

    SciTech Connect (OSTI)

    Van Blarigan, Peter; Shugard, Andrew D.; Walters, R. Tom (Savannah River National Labs, Aiken, SC)

    2012-01-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Mills, Bernice E.

    2007-11-01T23:59:59.000Z

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

  18. Information extraction system

    DOE Patents [OSTI]

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

    2014-05-13T23:59:59.000Z

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

  19. Facility Operations 1993 fiscal year work plan: WBS 1.3.1

    SciTech Connect (OSTI)

    Not Available

    1992-11-01T23:59:59.000Z

    The Facility Operations program is responsible for the safe, secure, and environmentally sound management of several former defense nuclear production facilities, and for the nuclear materials in those facilities. As the mission for Facility Operations plants has shifted from production to support of environmental restoration, each plant is making a transition to support the new mission. The facilities include: K Basins (N Reactor fuel storage); N Reactor; Plutonium-Uranium Reduction Extraction (PUREX) Plant; Uranium Oxide (UO{sub 3}) Plant; 300 Area Fuels Supply (N Reactor fuel supply); Plutonium Finishing Plant (PFP).

  20. Groundwater flow and tritium migration in coastal plain sediments, Savannah River Site, South Carolina

    SciTech Connect (OSTI)

    Harris, M.K. [Westinghouse Savannah River Company, Aiken, SC (United States); Flach, G.P.; Thayer, P.A. [Univ. of North Carolina (United States)

    1998-05-01T23:59:59.000Z

    Groundwater modeling was performed to assess groundwater flow and contaminant migration for a tritium plume at the Savannah River Site (SRS). The study supports the Corrective Measures Study and Interim Action Plan regulatory documents for the Old Radioactive Waste Burial Ground (ORWBG). Modeling scenarios were designed to provide data for an economic analysis of alternatives, and subsequently evaluate the effectiveness of the selected remedial technologies for tritium reduction to surface waters. Scenarios assessed include no action, vertical and surface barriers, pump-treat-reinject, and vertical recirculation wells. Hydrostratigraphic units in the area consist of fluvial, deltaic, and shallow marine sand, mud, and calcareous sediments that exhibit abrupt facies changes over short distances. The complex heterogeneity of the sediments, along with characterization data, and tritium contaminant source data required a three-dimensional model be developed in order to accurately illustrate the size, shape and orientation of the plume. Results demonstrate that the shallow confining zone in the region controls the migration path of the plume. The size and shape of the plume were modeled in three-dimensions using detailed core, geophysical and cone-penetrometer data, depth-discrete contaminant data, monitoring well data, and seepline/surface water samples. Three-dimensional tritium plume maps were created for the >20,000, >500 and >50 pCi/ml concentration levels. The three-dimensional plume maps and volumetric calculations indicate that 63 percent of the total activity and 12 percent of the volume above 50 pCi/ml resides in a layer less than 6-m thick riding on top of the shallow confining zone.

  1. Determination of transport parameters from coincident chloride and tritium plumes at the Idaho National Engineering Laboratory

    E-Print Network [OSTI]

    Fryar, Alan Ernest

    1986-01-01T23:59:59.000Z

    -radioactive waste, but rad1onuclides are often toxic at far lower concentrations than are hazardous non-radi oacti ve speci es (Freeze and Cherry, 1979). Most radioactive waste, in terms of activity, is generated at vari ous stages of what Freeze and Cherry...DETERMINATION OF TRANSPORT PARAMETERS FROM COINCIDENT CHLORIDE AND TRITIUM PLUMES AT THE IDAHO NATIONAL ENGINEERING LABORATORY A Thesis by ALAN ERNEST FRYAR Submitted to the Graduate College of Texas A&M University in partial fulfillment...

  2. Intercomparison run for uranium and tritium determination in urine samples, organised by Nuclear Regulatory Authority, Argentina

    E-Print Network [OSTI]

    Serdeiro, N H; Equillor, H E

    2003-01-01T23:59:59.000Z

    The Nuclear Regulatory Authority (ARN), Argentina, has carried out an intercomparison run for tritium and uranium determination in urine, in November 2002. The aim of this exercise was to assess the performance of the laboratories that usually inform these radionuclides and to provide technical support in order to have an appropriate occupational monitoring in vitro. In the present work, the results of the intercomparison and the assessment of each laboratory are published.

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

    SciTech Connect (OSTI)

    Carsten, A.L.

    1984-01-01T23:59:59.000Z

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

  4. Tritium and helium analyses in thin films by enhanced proton backscattering

    E-Print Network [OSTI]

    Tao Fu; Zhu An; Jing-Jun Zhu; Man-Tian Liu; Li Mao

    2013-10-14T23:59:59.000Z

    In order to perform quantitative tritium and helium analysis in thin film sample by using enhanced proton backscattering (EPBS), EPBS spectra for several samples consisting of non-RBS light elements (i.e., T, 4He, 12C, 16O, natSi), medium and heavy elements have been measured and analyzed by using analytical SIMNRA and Monte Carlo-based CORTEO codes. The CORTEO code used in this paper is modified and some non-RBS cross sections of proton scattering from T, 4He, 12C, 14N, 16O and natSi elements taken from ENDF/B-VII.1 database and the calculations of SigmaCalc code are incorporated. All cross section data needed in CORTEO code over the entire proton incident energy-scattering angle plane are obtained by interpolation. It is quantitatively observed that the multiple and plural scattering effects have little impact on energy spectra for light elements like T, He, C, O and Si, and the RBS cross sections of light elements, instead of the non-RBS cross sections, can be used in SIMNRA code for dual scattering calculations for EPBS analysis. It is also observed that at the low energy part of energy spectrum the results given by CORTEO code are higher than the results of SIMNRA code and are in better agreement with the experimental data, especially when heavier elements exist in samples. For tritium analysis, the tritium depth distributions should not be simply adjusted to fit the experimental spectra when the multiple and plural scattering contributions are not completely accounted, or else inaccurate results may be obtained. For medium and heavy matrix elements, when full Monte Carlo RBS calculations are used in CORTEO code, the results from CORTEO code are in good agreement with the experimental results at the low energy part of energy spectra, at this moment quantitative tritium and helium analysis in thin film sample by using enhanced proton backscattering can be performed reliably.

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

    DOE Patents [OSTI]

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

    1985-01-01T23:59:59.000Z

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

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

    DOE Patents [OSTI]

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

    1983-06-10T23:59:59.000Z

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

  7. ACHIEVING THE REQUIRED COOLANT FLOW DISTRIBUTION FOR THE ACCELERATOR PRODUCTION OF TRITIUM (APT) TUNGSTEN NEUTRON SOURCE

    SciTech Connect (OSTI)

    D. SIEBE; K. PASAMEHMETOGLU

    2000-11-01T23:59:59.000Z

    The Accelerator Production of Tritium neutron source consists of clad tungsten targets, which are concentric cylinders with a center rod. These targets are arranged in a matrix of tubes, producing a large number of parallel coolant paths. The coolant flow required to meet thermal-hydraulic design criteria varies with location. This paper describes the work performed to ensure an adequate coolant flow for each target for normal operation and residual heat-removal conditions.

  8. Facility effluent monitoring plan determinations for the 300 Area facilities

    SciTech Connect (OSTI)

    Nickels, J.M.

    1991-08-01T23:59:59.000Z

    Facility Effluent Monitoring Plan determinations were conducted for the Westinghouse Hanford Company 300 Area facilities on the Hanford Site. These determinations have been prepared in accordance with A Guide For Preparing Hanford Site Facility Effluent Monitoring Plans. Sixteen Westinghouse Hanford Company facilities in the 300 Area were evaluated: 303 (A, B, C, E, F, G, J and K), 303 M, 306 E, 308, 309, 313, 333, 334 A, and the 340 Waste Handling Facility. The 303, 306, 313, 333, and 334 facilities Facility Effluent Monitoring Plan determinations were prepared by Columbia Energy and Environmental Services of Richland, Washington. The 340 Central Waste Complex determination was prepared by Bovay Northwest, Incorporated. The 308 and 309 facility determinations were prepared by Westinghouse Handford Company. Of the 16 facilities evaluated, 3 will require preparation of a Facility effluent Monitoring Plan: the 313 N Fuels Fabrication Support Building, 333 N Fuels fabrication Building, and the 340 Waste Handling Facility. 26 refs., 5 figs., 10 tabs.

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

    SciTech Connect (OSTI)

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

    1990-01-01T23:59:59.000Z

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

  10. Search for an admixture of sterile neutrino in the electron spectrum from tritium $?$-decay

    E-Print Network [OSTI]

    D. Abdurashitov; A. Berlev; N. Likhovid; A. Lokhov; I. Tkachev; V. Yants

    2014-05-15T23:59:59.000Z

    We propose an experiment intended for search for an admixture of sterile neutrino with mass m$_s$ in the range of 1-8 keV that may be detected as specific distortion of the electron energy spectrum during tritium decay. The distortion is spread over large part of the spectrum so to reveal it one can use a detector with relatively poor (near 10-15%) energy resolution. A classic proportional counter is a simple natural choice for a tritium $\\beta$-decay detector. The method we are proposing is original in two respects. First, the counter is produced as a whole from fully-fused quartz tube allowing to measure current pulse directly from anode while providing high stability for a long time. Second, a modern digital acquisition technique can be used in measurements at ultrahigh count rate - up to 10$^6$ Hz. As a result an energy spectrum of tritium electrons containing up to 10$^{12}$ counts may be collected in a month of live time measurements. Due to high statistics an upper limit down to 10$^{-3}$..10$^{-5}$ can be put on sterile neutrino mixing at 95% CL for m$_s$ in the range of 1-8 keV, that will be 1..2 orders of magnitude better then bounds published up to now.

  11. Preliminary risks associated with postulated tritium release from production reactor operation

    SciTech Connect (OSTI)

    O'Kula, K.R.; Horton, W.H.

    1988-01-01T23:59:59.000Z

    The Probabilistic Risk Assessment (PRA) of Savannah River Plant (SRP) reactor operation is assessing the off-site risk due to tritium releases during postulated full or partial loss of heavy water moderator accidents. Other sources of tritium in the reactor are less likely to contribute to off-site risk in non-fuel melting accident scenarios. Preliminary determination of the frequency of average partial moderator loss (including incidents with leaks as small as .5 kg) yields an estimate of /approximately/1 per reactor year. The full moderator loss frequency is conservatively chosen as 5 /times/ 10/sup /minus/3/ per reactor year. Conditional consequences, determined with a version of the MACCS code modified to handle tritium, are found to be insignificant. The 95th percentile individual cancer risk is 4 /times/ 10/sup /minus/8/ per reactor year within 16 km of the release point. The full moderator loss accident contributes about 75% of the evaluated risks. 13 refs., 4 figs., 5 tabs.

  12. Nuclear Power Generating Facilities (Maine)

    Broader source: Energy.gov [DOE]

    The first subchapter of the statute concerning Nuclear Power Generating Facilities provides for direct citizen participation in the decision to construct any nuclear power generating facility in...

  13. Pollution Control Facilities (South Carolina)

    Broader source: Energy.gov [DOE]

    For the purpose of this legislation, pollution control facilities are defined as any facilities designed for the elimination, mitigation or prevention of air or water pollution, including all...

  14. LANL | Physics | Trident Laser Facility

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

    Discovery science at Trident Laser Facility Several important discoveries and first observations have been made at the Trident Laser Facility, a unique three-beam neodymium-glass...

  15. Hazardous Waste Facilities Siting (Connecticut)

    Broader source: Energy.gov [DOE]

    These regulations describe the siting and permitting process for hazardous waste facilities and reference rules for construction, operation, closure, and post-closure of these facilities.

  16. Sandia National Laboratories: SWIFT Facility

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

    SWIFT Facility Characterizing Scaled Wind Farm Technology Facility Inflow On April 1, 2014, in Energy, News, News & Events, Partnership, Renewable Energy, Wind Energy The Scaled...

  17. User Facilities | Argonne National Laboratory

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

    User Facilities Advanced Photon Source Argonne Leadership Computing Facility Argonne Tandem Linear Accelerator System Center for Nanoscale Materials Transportation Research and...

  18. Cornell University Facilities Services

    E-Print Network [OSTI]

    Manning, Sturt

    Description: The Large Animal Teaching Complex (LATC) will be a joint facility for the College of Veterinary or increase operating costs of the dairy barn; therefore, the College of Veterinary Medicine has agreed

  19. Photovoltaic Research Facilities

    Broader source: Energy.gov [DOE]

    The U.S. Department of Energy (DOE) funds photovoltaic (PV) research and development (R&D) at its national laboratory facilities located throughout the country. To encourage further innovation,...

  20. NEW RENEWABLE FACILITIES PROGRAM

    E-Print Network [OSTI]

    's electricity from renewable resources by 2010. The Guidebook outlines eligibility and legal requirementsCALIFORNIA ENERGY COMMISSION ` NEW RENEWABLE FACILITIES PROGRAM GUIDEBOOK March 2007 CEC-300 Executive Director Heather Raitt Technical Director RENEWABLE ENERGY OFFICE CALIFORNIA ENERGY COMMISSION

  1. NETL - Fuel Reforming Facilities

    SciTech Connect (OSTI)

    None

    2013-06-12T23:59:59.000Z

    Research using NETL's Fuel Reforming Facilities explores catalytic issues inherent in fossil-energy related applications, including catalyst synthesis and characterization, reaction kinetics, catalyst activity and selectivity, catalyst deactivation, and stability.

  2. NETL - Fuel Reforming Facilities

    ScienceCinema (OSTI)

    None

    2014-06-27T23:59:59.000Z

    Research using NETL's Fuel Reforming Facilities explores catalytic issues inherent in fossil-energy related applications, including catalyst synthesis and characterization, reaction kinetics, catalyst activity and selectivity, catalyst deactivation, and stability.

  3. Liquidity facilities and signaling

    E-Print Network [OSTI]

    Arregui, Nicolás

    2010-01-01T23:59:59.000Z

    This dissertation studies the role of signaling concerns in discouraging access to liquidity facilities like the IMF contingent credit lines (CCL) and the Discount Window (DW). In Chapter 1, I analyze the introduction of ...

  4. Facilities Management Department Restructuring

    E-Print Network [OSTI]

    Mullins, Dyche

    ­ Zone 2 ­ Mission Bay/East Side: Includes Mission Bay, Mission Center Bldg, Buchanan Dental, Hunters Point, 654 Minnesota, Oyster Point 2. Recommendation that UCSF align all Facility Services and O

  5. Sandia National Laboratories: Facilities

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

    Renewable Energy, SWIFT, Wind Energy One of the primary roles of Sandia's Scaled Wind Farm Technology (SWiFT) facility will be to conduct detailed experiments on turbine wakes...

  6. TUDE DES CIBLES MINCES CHARGES DE TRITIUM UTILISES DANS LES ACCLRATEURS PRODUISANT DES NEUTRONS RAPIDES PAR LA RACTION D-T

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    61 �TUDE DES CIBLES MINCES CHARG�ES DE TRITIUM UTILIS�ES DANS LES ACC�L�RATEURS PRODUISANT DES [titane-tritium]préparées par le Département des Radioéléments de Saclay. Dix sept cibles ont été ainsi rendement. Abstract. 2014 This study permits the determination of the efficiency of titan-tritium targets

  7. la cellule en cuivre en contact avec l'hydrogne liquide (fig. 3). La liaison entre la cellule et le rservoir tritium

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    et le réservoir tritium gazeux est un tube d'acier inoxydable de 2 mm de diamètre. Les essais au tritium, dont le point triple est de 20,3 ~K (communiqué par J. K. Seagrave). Afin d'éviter une solidification dangereuse du tritium, une pressurisation de 0 à 100 g~cm2 environ est prévue sur le cryostat à

  8. Hanford Site Near-Facility Environmental Monitoring Data Report for Calendar Year 2008

    SciTech Connect (OSTI)

    Perkins, Craig J.; Dorsey, Michael C.; Mckinney, Stephen M.; Wilde, Justin W.; Poston, Ted M.

    2009-09-15T23:59:59.000Z

    Near-facility environmental monitoring is defined as monitoring near facilities that have the potential to discharge or have discharged, stored, or disposed of radioactive or hazardous materials. Monitoring locations are associated with nuclear facilities such as the Plutonium Finishing Plant, Canister Storage Building, and the K Basins; inactive nuclear facilities such as N Reactor and the Plutonium-Uranium Extraction (PUREX) Facility; and waste storage or disposal facilities such as burial grounds, cribs, ditches, ponds, tank farms, and trenches. Much of the monitoring consists of collecting and analyzing environmental samples and methodically surveying areas near facilities. The program is also designed to evaluate acquired analytical data, determine the effectiveness of facility effluent monitoring and controls, assess the adequacy of containment at waste disposal units, and detect and monitor unusual conditions.

  9. Strategies for Facilities Renewal

    E-Print Network [OSTI]

    Good, R. L.

    psig * Plant or Service Air 90 psig * Starting Air for gas engines 220 psig * Instrument Air 80 psig * 02 - process * N2 high purity 4. Water production systems and distribution * Potable water (remote rural site) * Fire water (not treated) * Cooling... sewers 6. Fuel systems * Mixed fuel (both by-product and purchased methane) * Pipeline natural gas * Fuel oil 7. Maintenance and office facilities * Various maintenance/construction shops, stores, offices * Office facilities for technical...

  10. Mound facility physical characterization

    SciTech Connect (OSTI)

    Tonne, W.R.; Alexander, B.M.; Cage, M.R.; Hase, E.H.; Schmidt, M.J.; Schneider, J.E.; Slusher, W.; Todd, J.E.

    1993-12-01T23:59:59.000Z

    The purpose of this report is to provide a baseline physical characterization of Mound`s facilities as of September 1993. The baseline characterizations are to be used in the development of long-term future use strategy development for the Mound site. This document describes the current missions and alternative future use scenarios for each building. Current mission descriptions cover facility capabilities, physical resources required to support operations, current safety envelope and current status of facilities. Future use scenarios identify potential alternative future uses, facility modifications required for likely use, facility modifications of other uses, changes to safety envelope for the likely use, cleanup criteria for each future use scenario, and disposition of surplus equipment. This Introductory Chapter includes an Executive Summary that contains narrative on the Functional Unit Material Condition, Current Facility Status, Listing of Buildings, Space Plans, Summary of Maintenance Program and Repair Backlog, Environmental Restoration, and Decontamination and Decommissioning Programs. Under Section B, Site Description, is a brief listing of the Site PS Development, as well as Current Utility Sources. Section C contains Site Assumptions. A Maintenance Program Overview, as well as Current Deficiencies, is contained within the Maintenance Program Chapter.

  11. 2-D Spatial Distribution of D-D and D-T Neutron Emission in JET ELMy H-mode Plasmas with Tritium Puff

    E-Print Network [OSTI]

    2-D Spatial Distribution of D-D and D-T Neutron Emission in JET ELMy H-mode Plasmas with Tritium Puff

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

    E-Print Network [OSTI]

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

  13. The Preparative Gas Chromatographic System of the JET Active Gas Handling System ­ Tritium Commissioning and use during and after DTE1

    E-Print Network [OSTI]

    The Preparative Gas Chromatographic System of the JET Active Gas Handling System ­ Tritium Commissioning and use during and after DTE1

  14. Regional groundwater flow and tritium transport modeling and risk assessment of the underground test area, Nevada Test Site, Nevada

    SciTech Connect (OSTI)

    None

    1997-10-01T23:59:59.000Z

    The groundwater flow system of the Nevada Test Site and surrounding region was evaluated to estimate the highest potential current and near-term risk to the public and the environment from groundwater contamination downgradient of the underground nuclear testing areas. The highest, or greatest, potential risk is estimated by assuming that several unusually rapid transport pathways as well as public and environmental exposures all occur simultaneously. These conservative assumptions may cause risks to be significantly overestimated. However, such a deliberate, conservative approach ensures that public health and environmental risks are not underestimated and allows prioritization of future work to minimize potential risks. Historical underground nuclear testing activities, particularly detonations near or below the water table, have contaminated groundwater near testing locations with radioactive and nonradioactive constituents. Tritium was selected as the contaminant of primary concern for this phase of the project because it is abundant, highly mobile, and represents the most significant contributor to the potential radiation dose to humans for the short term. It was also assumed that the predicted risk to human health and the environment from tritium exposure would reasonably represent the risk from other, less mobile radionuclides within the same time frame. Other contaminants will be investigated at a later date. Existing and newly collected hydrogeologic data were compiled for a large area of southern Nevada and California, encompassing the Nevada Test Site regional groundwater flow system. These data were used to develop numerical groundwater flow and tritium transport models for use in the prediction of tritium concentrations at hypothetical human and ecological receptor locations for a 200-year time frame. A numerical, steady-state regional groundwater flow model was developed to serve as the basis for the prediction of the movement of tritium from the underground testing areas on a regional scale. The groundwater flow model was used in conjunction with a particle-tracking code to define the pathlines followed by groundwater particles originating from 415 points associated with 253 nuclear test locations. Three of the most rapid pathlines were selected for transport simulations. These pathlines are associated with three nuclear test locations, each representing one of the three largest testing areas. These testing locations are: BOURBON on Yucca Flat, HOUSTON on Central Pahute Mesa, and TYBO on Western Pahute Mesa. One-dimensional stochastic tritium transport simulations were performed for the three pathlines using the Monte Carlo method with Latin hypercube sampling. For the BOURBON and TYBO pathlines, sources of tritium from other tests located along the same pathline were included in the simulations. Sensitivity analyses were also performed on the transport model to evaluate the uncertainties associated with the geologic model, the rates of groundwater flow, the tritium source, and the transport parameters. Tritium concentration predictions were found to be mostly sensitive to the regional geology in controlling the horizontal and vertical position of transport pathways. The simulated concentrations are also sensitive to matrix diffusion, an important mechanism governing the migration of tritium in fractured carbonate and volcanic rocks. Source term concentration uncertainty is most important near the test locations and decreases in importance as the travel distance increases. The uncertainty on groundwater flow rates is as important as that on matrix diffusion at downgradient locations. The risk assessment was performed to provide conservative and bounding estimates of the potential risks to human health and the environment from tritium in groundwater. Risk models were designed by coupling scenario-specific tritium intake with tritium dose models and cancer and genetic risk estimates using the Monte Carlo method. Estimated radiation doses received by individuals from chronic exposure to tritium, and the corre

  15. THE RADIOLOGICAL RESEARCH ACCELERATOR FACILITY The Radiological Research Accelerator Facility

    E-Print Network [OSTI]

    THE RADIOLOGICAL RESEARCH ACCELERATOR FACILITY 71 The Radiological Research Accelerator Facility the irradiated cells. Both the microbeam and the track segment facilities continue to be utilized in various investigations of this phenomenon. The single- particle microbeam facility provides precise control of the number

  16. THE RADIOLOGICAL RESEARCH ACCELERATOR FACILITY The Radiological Research Accelerator Facility

    E-Print Network [OSTI]

    THE RADIOLOGICAL RESEARCH ACCELERATOR FACILITY 1 The Radiological Research Accelerator Facility for Radiological Research (CRR). Using the mi- crobeam facility, 10% of the cells were irradiated through particle beam as well as the first fo- cused microbeam in the new microbeam facility. · Another significant

  17. Facility Location with Hierarchical Facility Costs Zoya Svitkina #

    E-Print Network [OSTI]

    Tardos, Ă?va

    Facility Location with Hierarchical Facility Costs Zoya Svitkina # â?? Eva Tardos + Abstract We consider the facility location problem with hierarchi­ cal facility costs, and give a (4 installation costs. Shmoys, Swamy and Levi [13] gave an approxi­ mation algorithm for a two­level version

  18. National Ignition Facility Project Completion and Control System Status

    SciTech Connect (OSTI)

    Van Arsdall, P J; Azevedo, S G; Beeler, R G; Bryant, R M; Carey, R W; Demaret, R D; Fisher, J M; Frazier, T M; Lagin, L J; Ludwigsen, A P; Marshall, C D; Mathisen, D G; Reed, R K

    2009-10-02T23:59:59.000Z

    The National Ignition Facility (NIF) is the world's largest and most energetic laser experimental system providing a scientific center to study inertial confinement fusion (ICF) and matter at extreme energy densities and pressures. Completed in 2009, NIF is a stadium-sized facility containing a 1.8-MJ, 500-TW 192-beam ultraviolet laser and target chamber. A cryogenic tritium target system and suite of optical, X-ray and nuclear diagnostics will support experiments in a strategy to achieve fusion ignition starting in 2010. Automatic control of NIF is performed by the large-scale Integrated Computer Control System (ICCS), which is implemented by 2 MSLOC of Java and Ada running on 1300 front-end processors and servers. The ICCS framework uses CORBA distribution for interoperation between heterogeneous languages and computers. Laser setup is guided by a physics model and shots are coordinated by data-driven distributed workflow engines. The NIF information system includes operational tools and a peta-scale repository for provisioning experimental results. This paper discusses results achieved and the effort now underway to conduct full-scale operations and prepare for ignition.

  19. NREL: Energy Systems Integration Facility - Facility Design

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office of Science (SC)Integrated CodesTransparency Visit | NationalWebmaster ToStaffCapabilities TheFacility

  20. Radiation Effects Facility - Facilities - Cyclotron Institute

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office of Science (SC)IntegratedSpeedingTechnicalPurchase, Delivery, andSmartRadiation Effects Facility

  1. UNIVERSITY BOULEVARD FAU Research Facility

    E-Print Network [OSTI]

    Fernandez, Eduardo

    Harriet L.Wilkes Honors College FAU Research Facility Expansion Satellite Utility Plant Chiller Lift

  2. Fission Product Extraction Process

    ScienceCinema (OSTI)

    None

    2013-05-28T23:59:59.000Z

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

  3. PUREX/UO{sub 3} facilities deactivation lessons learned history

    SciTech Connect (OSTI)

    Hamrick, D.G.; Gerber, M.S.

    1995-01-01T23:59:59.000Z

    The Plutonium-Uranium Extraction (PUREX) Facility operated from 1956-1972, from 1983-1988, and briefly during 1989-1990 to produce for national defense at the Hanford Site in Washington State. The Uranium Trioxide (UO{sub 3}) Facility operated at the Hanford Site from 1952-1972, 1984-1988, and briefly in 1993. Both plants were ordered to permanent shutdown by the U.S. Department of Energy (DOE) in December 1992, thus initiating their deactivation phase. Deactivation is that portion of a facility`s life cycle that occurs between operations and final decontamination and decommissioning (D&D). This document details the history of events, and the lessons learned, from the time of the PUREX Stabilization Campaign in 1989-1990, through the end of the first full fiscal year (FY) of the deactivation project (September 30, 1994).

  4. ARM - Facility News Article

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office of ScienceandMesa del(ANL-IN-03-032)8Li (59AJ76)ARM2, 2006 [Facility News]SPARTICUSJune31, 2005 [Facility

  5. Facilities | Department of Energy

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011 Strategic Plan| Department of.pdf6-OPAMDepartment6 FY 2007FY 2014Facilities Facilities

  6. Facility Disposition Projects

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011 Strategic Plan| Department of.pdf6-OPAMDepartment6 FY 2007FY 2014Facilities Facilities

  7. Facility Data Policy

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office of Science (SC) Environmental Assessments (EA)Budget » FY 2014Facilities FusionFacility Data Policy

  8. 5.10 Tritium Geochemistry And Kd Values 5.10.1 Overview: Important Aqueous-and Solid-Phase Parameters

    E-Print Network [OSTI]

    mainly by the nuclear interaction of nitrogen with fast neutrons induced by cosmic ray reactions, respectively (Freeze and Cherry, 1979). Tritium can also be created in nuclear reactors as a result NRC Site Decommissioning Site Plan (SDMP) sites. 5.10.3 Aqueous Speciation Because tritium oxidizes

  9. Sensitivity of Next-Generation Tritium Beta-Decay Experiments for keV-Scale Sterile Neutrinos

    E-Print Network [OSTI]

    S. Mertens; T. Lasserre; S. Groh; F. Glueck; A. Huber; A. W. P. Poon; M. Steidl; N. Steinbrink; C. Weinheimer

    2014-12-14T23:59:59.000Z

    We investigate the sensitivity of tritium $\\beta$-decay experiments for keV-scale sterile neutrinos. Relic sterile neutrinos in the keV mass range can contribute both to the cold and warm dark matter content of the universe. This work shows that a large-scale tritium beta-decay experiment, similar to the KATRIN experiment that is under construction, can reach a statistical sensitivity of the active-sterile neutrino mixing of $\\sin^2\\theta \\sim 10^{-8}$. The effect of uncertainties in the known theoretical corrections to the tritium $\\beta$-decay spectrum were investigated, and found not to affect the sensitivity significantly. It is demonstrated that controlling uncorrelated systematic effects will be one of the main challenges in such an experiment.

  10. FACILITIES INSTRUCTIONS, STANDARDS, & TECHNIQUES

    E-Print Network [OSTI]

    Laughlin, Robert B.

    to the repair of hydraulic turbine runners and large pump impellers. Reclamation operates and maintains a wideFACILITIES INSTRUCTIONS, STANDARDS, & TECHNIQUES VOLUME 2-5 TURBINE REPAIR Internet Version variety of reaction and impulse turbines as well as axial flow, mixed flow, radial flow pumps and pump

  11. Facilities Management Field Services

    E-Print Network [OSTI]

    Hickman, Mark

    Facilities Management Field Services FieldStationsAnnualReport2006 #12;Cover Photo by Dr Mark Jermy coast #12; Introduction A very wet Steve Weaver emerges from the river. Ah, field work! The Government broadband, at least there is now an alternative to the telephone line. Electrical power spikes (and outages

  12. Graph algorithms experimentation facility

    E-Print Network [OSTI]

    Sonom, Donald George

    1994-01-01T23:59:59.000Z

    DRAWADJMAT 2 ~e ~l 2. ~f ~2 2 ~t ~& [g H 2 O? Z Mwd a P d ed d Aid~a sae R 2-BE& T C dbms Fig. 2. External Algorithm Handler The facility is menu driven and implemented as a client to XAGE. Our implementation follows very closely the functionality...

  13. NEW RENEWABLE FACILITIES PROGRAM

    E-Print Network [OSTI]

    for and receive production incentives, referred to as supplemental energy payments (SEPs), from the New RenewableCALIFORNIA ENERGY COMMISSION NEW RENEWABLE FACILITIES PROGRAM GUIDEBOOK APRIL 2006 CEC-300 Director Heather Raitt Technical Director Renewable Energy Program Drake Johnson Office Manager Renewable

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

    SciTech Connect (OSTI)

    Subramanian, K; Michael Morgan, M

    2004-01-10T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    None

    1996-12-01T23:59:59.000Z

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

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

    E-Print Network [OSTI]

    Rice, Brent Lee

    1987-01-01T23:59:59.000Z

    power increases with time. The subscript 1 and 2 utilized in Fig. 1 refer to the fusile and fissile fuels, respectively. The subscripts H, F, and T refer to the fusion reactor, the fission power Fissile Trrtium (1-a) (CH CR1) RH Frssion Power... and fission power reactors, respectively, are: 13 d NT2 (C 1 ) R + a (C C 1 ) R dt (7) d NF2 (C -1) R + (1-a) (C -C ) R dt where N denotes the fuel inventory of a particular system component. C&1 and CT1 denote the number of tritium atoms produced...

  17. Exact relativistic tritium beta-decay endpoint spectrum in a hadron model

    E-Print Network [OSTI]

    Fedor Simkovic; Rastislav Dvornicky; Amand Faessler

    2008-05-05T23:59:59.000Z

    We present the relativistic calculation of the beta-decay of tritium in a hadron model. The elementary particle treatment of the transition 3H -> 3He + e^- + nu_e is performed in analogy with the description of the beta-decay of neutron. The effects of higher order terms of hadron current and nuclear recoil are taken into account in this formalism. The relativistic Kurie function is derived and presented in a simple form suitable for the determination of neutrino masses from the shape of the endpoint spectrum. A connection with the commonly used Kurie function is established.

  18. Recoilless Resonance Absorption of Tritium Antineutrinos and Time-Energy Uncertainty Relation

    E-Print Network [OSTI]

    S. M. Bilenky

    2007-08-02T23:59:59.000Z

    We discuss neutrino oscillations in an experiment with M\\"ossbauer recoilless resonance absorbtion of tritium antineutrinos, proposed recently by Raghavan. We demonstrate that small energy uncertainty of antineutrinos which ensures a large resonance absorption cross section is in a conflict with the energy uncertainty which, according to the time-energy uncertainty relation, is necessary for neutrino oscillations to happen. The search for neutrino oscillations in the M\\"ossbauer neutrino experiment would be an important test of the applicability of the time-energy uncertainty relation to a newly discovered interference phenomenon.

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

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels DataDepartment of Energy Your Density Isn't Your Destiny:Revised Finding of98-F,-SA-01: Supplement AnalysisDepartment of(CLWR) Tritium

  20. Plasma Fueling, Pumping, and Tritium Handling Considerations for FIRE P.W. Fisher', M. J.Gouge', C. A. Foster',B. E. Nelson', C. A. Gentile' andthe FIRE StudyTeam

    E-Print Network [OSTI]

    Plasma Fueling, Pumping, and Tritium Handling Considerations for FIRE P.W. Fisher', M. J.Gouge', C,P.O.Box 2009,OakRidge,TN 3783l-8071 *PrincetonPlasmaPhysicsLaboratory, P.O.Box 451,Princeton,NJ 08543 Abstract-Tritium pellet injection will be utilized on the Fusion Ignition Research Experiment (FIRE) for efficient tritium

  1. Results of the quarterly tritium survey of Fourmile Branch and its seeplines in the F- and H-Areas of SRS: September 1993

    SciTech Connect (OSTI)

    Dixon, K.L.; Rogers, V.A.; Looney, B.B.

    1994-06-01T23:59:59.000Z

    The Environmental Sciences Section (ESS) established a quarterly monitoring program of the Fourmile Branch (FMB) seepline down gradient from the F- and H-Area seepage basins. The program surveys and tracks changes in tritium, specific conductivity, and pH for the seepline water. Measurements from the sixth quarterly survey (September 1993) showed higher tritium and conductivity measurements and higher pH values (pH 5 - 6) than measurements from previous studies. Increased tritium concentrations and conductivity values, as compared to previous surveys, were attributed to decreased rainfall prior to the sampling event However, overall results of the tritium survey and stream monitoring data (Looney et al., 1993) suggest that the tritium plume is flushing from the FMB system.

  2. Nano Research Facility Lab Safety Manual Nano Research Facility

    E-Print Network [OSTI]

    Subramanian, Venkat

    1 Nano Research Facility Lab Safety Manual Nano Research Facility: Weining Wang Office: Brauer---chemical, biological, or radiological. Notify the lab manager, Dr. Yujie Xiong at 5-4530. Eye Contact: Promptly flush

  3. Biomass Anaerobic Digestion Facilities and Biomass Gasification Facilities (Indiana)

    Broader source: Energy.gov [DOE]

    The Indiana Department of Environmental Management requires permits before the construction or expansion of biomass anaerobic digestion or gasification facilities.

  4. Evaluation of the tritium content in light water reactor control and absorber rods to obtain data for the fuel cycle backend

    SciTech Connect (OSTI)

    Bleier, A.; Neeb, K.H.; Gelfort, E.; Mischke, J.

    1986-08-01T23:59:59.000Z

    Tritium inventories and tritium distribution have been determined in boron glass absorber rods discharged from a pressurized water reactor first-cycle core and in spent boron carbide (B/sub 4/C) control rods from a boiling water reactor. The total tritium inventory in the boron glass absorber rods from the Stade nuclear reactor amounts to approx. =8.0 x 10/sup 10/ Bq (2.2 Ci) per rod. Of this, 99.6% was fixed in the boron glass itself and 0.4% in the Al/sub 2/O/sub 3/ pellets. The 4 x 10/sup -3/% fractions in the tube cladding and support pipe and the 1 x 10/sup -2/% fraction in the fill gas accounted for an insignificant part of the total tritium inventory of the rod. This experimentally determined tritium inventory was a factor of 5 larger than that suggested by the calculated estimate. The discrepancy between analyzed and calculated values can be explained by tritium formation from lithium impurities in the boron glass, where a 30-ppm lithium content would be adequate for this tritium inventory to be generated by the reaction /sup 6/Li(n,..cap alpha..)/sup 3/H. Evaluation of the B/sub 4/C control rods from the Lingen nuclear reactor after 3 yr of operation gave a 3.2 x 10/sup 10/Bq(0.85-Ci)tritium inventory per B/sub 4/C rod, while the total tritium inventory for a control rod assembly containing 60 B/sub 4/C rods was approx. =1.9 x 10/sup 12/ Bq (50 Ci). The tritium generated was essentially bound 100% in the B/sub 4/C, since the hulls contained only 6 x 10/sup -3/% and the fill gas only 2 x 10/sup -4/%.

  5. PUREX/UO{sub 3} facilities deactivation lessons learned: History

    SciTech Connect (OSTI)

    Gerber, M.S.

    1997-11-25T23:59:59.000Z

    In May 1997, a historic deactivation project at the PUREX (Plutonium URanium EXtraction) facility at the Hanford Site in south-central Washington State concluded its activities (Figure ES-1). The project work was finished at $78 million under its original budget of $222.5 million, and 16 months ahead of schedule. Closely watched throughout the US Department of Energy (DOE) complex and by the US Department of Defense for the value of its lessons learned, the PUREX Deactivation Project has become the national model for the safe transition of contaminated facilities to shut down status.

  6. The Caterpillar Coal Gasification Facility 

    E-Print Network [OSTI]

    Welsh, J.; Coffeen, W. G., III

    1983-01-01T23:59:59.000Z

    This paper is a review of one of America's premier coal gasification installations. The caterpillar coal gasification facility located in York, Pennsylvania is an award winning facility. The plant was recognized as the 'pace setter plant of the year...

  7. Facilities Automation and Energy Management

    E-Print Network [OSTI]

    Jen, D. P.

    1983-01-01T23:59:59.000Z

    Computerized facilities automation and energy management systems can be used to maintain high levels of facilities operations efficiencies. The monitoring capabilities provides the current equipment and process status, and the analysis...

  8. Biomass Feedstock National User Facility

    Broader source: Energy.gov [DOE]

    Breakout Session 1B—Integration of Supply Chains I: Breaking Down Barriers Biomass Feedstock National User Facility Kevin L. Kenney, Director, Biomass Feedstock National User Facility, Idaho National Laboratory

  9. Analysis of the separation of protium from blanket tritium-product streams

    SciTech Connect (OSTI)

    Misra, B.; Maroni, V.A.

    1981-07-01T23:59:59.000Z

    The case is considered in which the blanket product stream has been purified to the point where only protium, tritium, and a small quantity of deuterium remain. A cryogenic distillation cascade concept developed specifically to handle this enrichment problem is shown. The concept is based on a series of distillation columns and equilibrators capable of producing a protium-rich stream containing less than 1000 appm T and a tritium-rich stream containing less than 2000 appm H. It is envisioned that both of these streams could be blended with streams of comparable composition in the mainstream position of the fuel cycle without further processing. The computational analysis of the cascade was based on a fixed arrangement of columns and equilibrators and a fixed number of theoretical plates per columns, since these features are less easily varied in an actual system than reflux ratios and flow rates. In order to test the flexibility of this conceptual enruchment system to adjust to variations of the H/T ratio in the feed, H/T values of 0.333, 1.00, and 3.00 were investigated.

  10. The Development of RF Heating of Magnetically Confined Deuterium-Tritium Plasmas

    SciTech Connect (OSTI)

    B.P. LeBlanc; C.K. Phillips; J.C. Hosea; R. Majeski; S. Bernabei [and others

    1999-06-01T23:59:59.000Z

    The experimental and theoretical development of ion cyclotron radiofrequency heating (ICRF) in toroidal magnetically-confined plasmas recently culminated with the demonstration of ICRF heating of D-T plasmas, first in the Tokamak Fusion Test Reactor (TFTR) and then in the Joint European Torus (JET). Various heating schemes based on the cyclotron resonances between the plasma ions and the applied ICRF waves have been used, including second harmonic tritium, minority deuterium, minority helium-3, mode conversion at the D-T ion-ion hybrid layer, and ion Bernstein wave heating. Second harmonic tritium heating was first shown to be effective in a reactor-grade plasma in TFTR. D-minority heating on JET has led to the achievement of Q = 0.22, the ratio of fusion power produced to RF power input, sustained over a few energy confinement times. In this paper, some of the key building blocks in the development of rf heating of plasmas are reviewed and prospects for the development of advanced methods of plasma control based on the application of rf waves are discussed.

  11. Reed Reactor Facility Annual Report

    SciTech Connect (OSTI)

    Frantz, Stephen G.

    2000-09-01T23:59:59.000Z

    This is the report of the operations, experiments, modifications, and other aspects of the Reed Reactor Facility for the year.

  12. CFTF | Carbon Fiber Technology Facility | ORNL

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

    BTRIC CNMS CSMB CFTF Working with CFTF HFIR MDF NTRC OLCF SNS Carbon Fiber Technology Facility Home | User Facilities | CFTF CFTF | Carbon Fiber Technology Facility SHARE Oak...

  13. CRAD, Nuclear Facility Construction - Structural Concrete, May...

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

    CRAD, Nuclear Facility Construction - Structural Concrete, May 29, 2009 CRAD, Nuclear Facility Construction - Structural Concrete, May 29, 2009 May 29, 2009 Nuclear Facility...

  14. THE RADIOLOGICAL RESEARCH ACCELERATOR FACILITY

    E-Print Network [OSTI]

    175 THE RADIOLOGICAL RESEARCH ACCELERATOR FACILITY #12;176 #12;177 THE RADIOLOGICAL RESEARCH the microbeam and the track-segment facilities have been utilized in various investigations. Table 1 lists-segment facility. Samples are treated with graded doses of radical scavengers to observe changes in the cluster

  15. Extraction Utility Design Specification

    Energy Savers [EERE]

    Extraction Utility Design Specification January 11, 2011 Document Version 1.9 1 Revision History Date Version Section and Titles Author Summary of Change January 15, 2010 1.0 All...

  16. Liquid chromatographic extraction medium

    DOE Patents [OSTI]

    Horwitz, E. Philip (Naperville, IL); Dietz, Mark L. (Evanston, IL)

    1994-01-01T23:59:59.000Z

    A method and apparatus for extracting strontium and technetium values from biological, industrial and environmental sample solutions using a chromatographic column is described. An extractant medium for the column is prepared by generating a solution of a diluent containing a Crown ether and dispersing the solution on a resin substrate material. The sample solution is highly acidic and is introduced directed to the chromatographic column and strontium or technetium is eluted using deionized water.

  17. Liquid chromatographic extraction medium

    DOE Patents [OSTI]

    Horwitz, E.P.; Dietz, M.L.

    1994-09-13T23:59:59.000Z

    A method and apparatus are disclosed for extracting strontium and technetium values from biological, industrial and environmental sample solutions using a chromatographic column. An extractant medium for the column is prepared by generating a solution of a diluent containing a Crown ether and dispersing the solution on a resin substrate material. The sample solution is highly acidic and is introduced directed to the chromatographic column and strontium or technetium is eluted using deionized water. 1 fig.

  18. Aspects of Cooling at the TRI$?$P Facility

    E-Print Network [OSTI]

    L. Willmann; G. P. Berg; U. Dammalapati; S. De; P. Dendooven; O. Dermois; K. Jungmann; A. Mol; C. J. G. Onderwater; A. Rogachevskiy; M. Sohani; E. Traykov; H. W. Wilschut

    2006-02-03T23:59:59.000Z

    The Tri$\\mu$P facility at KVI is dedicated to provide short lived radioactive isotopes at low kinetic energies to users. It comprised different cooling schemes for a variety of energy ranges, from GeV down to the neV scale. The isotopes are produced using beam of the AGOR cyclotron at KVI. They are separated from the primary beam by a magnetic separator. A crucial part of such a facility is the ability to stop and extract isotopes into a low energy beamline which guides them to the experiment. In particular we are investigating stopping in matter and buffer gases. After the extraction the isotopes can be stored in neutral atoms or ion traps for experiments. Our research includes precision studies of nuclear $\\beta$-decay through $\\beta$-$\

  19. Time-dependent tritium inventories and flow rates in fuel cycle components of a tokamak fusion reactor

    SciTech Connect (OSTI)

    Kuan, W.; Abdou, M.A. [Univ. of California, Los Angeles, CA (United States); Willms, R.S. [Los Alamos National Lab., NM (United States)

    1994-12-31T23:59:59.000Z

    The dynamic behavior of the fuel cycle in a fusion reactor is of crucial importance due to the need to keep track of the large amount of tritium being constantly produced, transported, and processed in the reactor system. Because tritium is a source of radioactivity, loss and exhaust to the environment must be kept to a minimum. With ITER advancing to its Engineering Design phase, there is a need to accurately predict the dynamic tritium inventories and flow rates throughout the fuel cycle and to study design variations to meet the demands of low tritium inventory. In this paper, time-dependent inventories and flow rates for several components of the fuel cycle are modeled and studied through the use of a new modular-type model for the dynamic simulation of the fuel cycle in a fusion reactor. The complex dynamic behavior in the modeled subsystems is analyzed using this new model. Previous dynamic models focusing on the fuel cycle dealt primarily with a residence time parameter ({tau}{sub res}) defining each subsystem of the model. In this modular model, this residence time approach is avoided in favor of a more accurate and flexible model that utilizes real design parameters and operating schedules of the various subsystems modeled.

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

    E-Print Network [OSTI]

    Abdou, Mohamed

    and TechnologyNuclear Science and Technology (FNST). 4 #12;FNST is the science, engineering, technology and materials Fusion Nuclear Science & Technology (FNST) FNST is the science, engineering, technologyBl k t T h l F l C l dBlanket Technology, Fuel Cycle and Tritium Self Sufficiency M h d Abd