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Note: This page contains sample records for the topic "tritium extraction facility" from the National Library of EnergyBeta (NLEBeta).
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1

Commercial Light Water Reactor Tritium Extraction Facility Geotechnical Summary Report  

SciTech Connect (OSTI)

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.

Lewis, M.R.

2000-01-11T23:59:59.000Z

2

Weapons engineering tritium facility overview  

SciTech Connect (OSTI)

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.

Najera, Larry [Los Alamos National Laboratory

2011-01-20T23:59:59.000Z

3

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

Broader source: Energy.gov [DOE]

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

4

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...

5

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...

6

An introduction to the National Tritium Labeling Facility  

SciTech Connect (OSTI)

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

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

1988-06-01T23:59:59.000Z

7

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

SciTech Connect (OSTI)

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

Not Available

1991-10-01T23:59:59.000Z

8

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

SciTech Connect (OSTI)

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.

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

1997-11-14T23:59:59.000Z

9

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

SciTech Connect (OSTI)

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.

Clark, E.A.

2003-01-07T23:59:59.000Z

10

Determination of tritiated formaldehyde in effluents from tritium facilities  

SciTech Connect (OSTI)

Recent observations suggested that formal-dehyde can be incorporated in vegetation at a very high rate. In this paper, the authors develop a methodology for determining tritiated formaldehyde (CHTO) in gaseous effluent containing HTO and HT as dominant species. CHTO being very soluble in water is collected in a solution of carrier formaldehyde. This carrier is necessary for precipitating for formaldehyde derivative of dimedone and collecting it by filtration. The precipitate, which contains the formaldehyde hydrogens, is freed from exchangeable tritium, dried in oven, and combusted to water for tritium determination. CHTO can thus be separated from HTO with a high efficiency, leading to the possibility of determining accurately 1 Bq of CHTO in as much as 5 {times} 10{sup 4} Bq of HTO. The methodology has been applied in preliminary experiments to determine the ratio of CHTO to HTO in effluent from a tritium-handling facility and effluent released from solid miscellaneous wastes.

Belot, Y.; Camus, H.; Marini, T. (Commissariat a l'Energie Atomique, DPEI/SERGD, BP 6, F-92265 Fontenay aux Roses Cedex (FR))

1992-03-01T23:59:59.000Z

11

Health physics manual of good practices for tritium facilities  

SciTech Connect (OSTI)

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.

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

1991-12-01T23:59:59.000Z

12

Radiological Characterization and Final Facility Status Report Tritium Research Laboratory  

SciTech Connect (OSTI)

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.

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

1996-08-01T23:59:59.000Z

13

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

SciTech Connect (OSTI)

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.

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

14

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.

15

NNSA TRITIUM SUPPLY CHAIN  

SciTech Connect (OSTI)

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.

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

16

Determination of a tritium bioassay technique for nuclear facilities  

E-Print Network [OSTI]

. Gaseous tritium is retained only to a small degree after inhalation, with over 905 released by exhaling; however, tritiated (') water (HTO) is almost completely retained. This degree of retention of tritiated water has been shown to occur... for ingestion, inhalation and transfer through the skin. Other comoounds of tritium will be re- tained to a greater or lesser extent dependent on the physiological use of the compound. Under normal condi- tions of fluid intake, the concentration of HTO...

Sensintaffar, Edwin Lee

1971-01-01T23:59:59.000Z

17

Summary of Topic1 Fusion Power Extraction  

E-Print Network [OSTI]

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

Abdou, Mohamed

18

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

SciTech Connect (OSTI)

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.

P. Calderoni; P. Sharpe; M. Shimada

2009-09-01T23:59:59.000Z

19

Tritium monitor  

DOE Patents [OSTI]

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.

Chastagner, P.

1994-06-14T23:59:59.000Z

20

Tritium monitor  

DOE Patents [OSTI]

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.

Chastagner, Philippe (Augusta, GA)

1994-01-01T23:59:59.000Z

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


21

Design of a deuterium and tritium-ablator shock ignition target for the National Ignition Facility  

SciTech Connect (OSTI)

Shock ignition presents a viable path to ignition and high gain on the National Ignition Facility (NIF). In this paper, we describe the development of the 1D design of 0.5 MJ class, all-deuterium and tritium (fuel and ablator) shock ignition target that should be reasonably robust to Rayleigh-Taylor fluid instabilities, mistiming, and hot electron preheat. The target assumes 'day one' NIF hardware and produces a yield of 31 MJ with reasonable allowances for laser backscatter, absorption efficiency, and polar drive power variation. The energetics of polar drive laser absorption require a beam configuration with half of the NIF quads dedicated to launching the ignitor shock, while the remaining quads drive the target compression. Hydrodynamic scaling of the target suggests that gains of 75 and yields 70 MJ may be possible.

Terry, Matthew R.; Perkins, L. John; Sepke, Scott M. [Lawrence Livermore National Laboratory, 7000 East Ave., Livermore, California 94550 (United States)

2012-11-15T23:59:59.000Z

22

Darlington tritium removal facility and station upgrading plant dynamic process simulation  

SciTech Connect (OSTI)

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

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

2008-07-15T23:59:59.000Z

23

Sources of tritium  

SciTech Connect (OSTI)

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

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

1980-12-01T23:59:59.000Z

24

Facility effluent monitoring plan for the Plutonium Uranium Extraction Facility  

SciTech Connect (OSTI)

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.

Greager, E.M.

1997-12-11T23:59:59.000Z

25

Facility effluent monitoring plan for the plutonium uranium extraction facility  

SciTech Connect (OSTI)

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.

Wiegand, D.L.

1994-09-01T23:59:59.000Z

26

Finding of no significant impact for the tritium facility modernization and consolidation project at the Savannah River Site  

SciTech Connect (OSTI)

The Department of Energy (DOE) has prepared an environmental assessment (EA) (DOE/EA-1222) for the proposed modernization and consolidation of the existing tritium facilities at the Savannah River Site (SRS), located near Aiken, South Carolina. Based on the analyses in the EA, DOE has determined that the proposed action is not a major Federal action significantly affecting the quality of the human environment within the meaning of the National Environmental Policy Act of 1969 (NEPA). Therefore, the preparation of an environmental impact statement (EIS) is not required, and DOE is issueing this Finding of No Significant Impact (FONSI).

NONE

1998-01-01T23:59:59.000Z

27

Accelerator Production of Tritium Programmatic Environmental Impact Statement Input Submittal  

SciTech Connect (OSTI)

The Programmatic Environmental Impact Statement for Tritium Supply and Recycling considers several methods for the production of tritium. One of these methods is the Accelerator Production of Tritium. This report summarizes the design characteristics of APT including the accelerator, target/blanket, tritium extraction facility, and the balance of plant. Two spallation targets are considered: (1) a tungsten neutron-source target and (2) a lead neutron-source target. In the tungsten target concept, the neutrons are captured by the circulating He-3, thus producing tritium; in the lead target concept, the tritium is produced by neutron capture by Li-6 in a surrounding lithium-aluminum blanket. This report also provides information to support the PEIS including construction and operational resource needs, waste generation, and potential routine and accidental releases of radioactive material. The focus of the report is on the impacts of a facility that will produce 3/8th of the baseline goal of tritium. However, some information is provided on the impacts of APT facilities that would produce smaller quantities.

Miller, L.A. [Sandia National Labs., Albuquerque, NM (United States); Greene, G.A. [Brookhaven National Lab., Upton, NY (United States); Boyack, B.E. [Los Alamos National Lab., NM (United States)

1996-02-01T23:59:59.000Z

28

Copper activation deuterium-tritium neutron yield measurements at the National Ignition Facility  

E-Print Network [OSTI]

, New Mexico 87131, USA 2 Sandia National Laboratories, Albuquerque, New Mexico 87185, USA 3 Lawrence Livermore National Laboratories, Livermore, California 94550, USA 4 Plasma Science and Fusion Center, MIT(+ ) and 65 Cu(n,2n) 64 Cu(+ ), has been fielded at the National Ignition Facility (NIF). The induced copper

29

TFTR tritium operations lessons learned  

SciTech Connect (OSTI)

The Tokamak Fusion Test Reactor which is the progenitor for full D-T operating tokamaks has successfully processed > 81 grams of tritium in a safe and efficient fashion. Many of the fundamental operational techniques associated with the safe movement of tritium through the TFTR facility were developed over the course of many years of DOE tritium facilities (LANL, LLNL, SRS, Mound). In the mid 1980`s The Tritium Systems Test Assembly (TSTA) at LANL began reporting operational techniques for the safe handling of tritium, and became a major conduit for the transfer of safe tritium handling technology from DOE weapons laboratories to non-weapon facilities. TFTR has built on many of the TSTA operational techniques and has had the opportunity of performing and enhancing these techniques at America`s first operational D-T fusion reactor. This paper will discuss negative pressure employing `elephant trunks` in the control and mitigation of tritium contamination at the TFTR facility, and the interaction between contaminated line operations and {Delta} pressure control. In addition the strategy employed in managing the movement of tritium through TFTR while maintaining an active tritium inventory of < 50,000 Ci will be discussed. 5 refs.

Gentile, C.A.; Raftopoulos, S.; LaMarche, P. [Princeton Plasma Physics Lab., NJ (United States)] [and others

1996-12-31T23:59:59.000Z

30

Performance of Vacuum Pumps to be Used in Tritium Extraction Facility  

SciTech Connect (OSTI)

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.

Steimke, J.L.

1999-04-06T23:59:59.000Z

31

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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarlyEnergyDepartmentNational NuclearhasAdministration77 SandiaGuidance to theJuly 2014 NMMSS NewsN I

32

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]

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

2006-01-01T23:59:59.000Z

33

Baseline radionuclide concentrations in soils and vegetation around the proposed Weapons Engineering Tritium Facility and the Weapons Subsystems Laboratory at TA-16  

SciTech Connect (OSTI)

A preoperational environmental survey is required by the Department of Energy (DOE) for all federally funded research facilities that have the potential to cause adverse impacts on the environment. Therefore, in accordance with DOE Order 5400.1, an environmental survey was conducted over the proposed sites of the Weapons Engineering Tritium Facility (WETF) and the Weapons Subsystems Laboratory (WSL) at Los Alamos National Laboratory (LANL) at TA-16. Baseline concentrations of tritium ({sup 3}H), plutonium ({sup 238}Pu and {sup 239}Pu) and total uranium were measured in soils, vegetation (pine needles and oak leaves) and ground litter. Tritium was also measured from air samples, while cesium ({sup 137}Cs) was measured in soils. The mean concentration of airborne tritiated water during 1987 was 3.9 pCi/m{sup 3}. Although the mean annual concentration of {sup 3}H in soil moisture at the 0--5 cm (2 in) soil depth was measured at 0.6 pCi/mL, a better background level, based on long-term regional data, was considered to be 2.6 pCi/mL. Mean values for {sup 137}Cs, {sup 218}Pu, {sup 239}Pu, and total uranium in soils collected from the 0--5 cm depth were 1.08 pCi/g, 0.0014 pCi/g, 0.0325 pCi/g, and 4.01 {micro}g/g, respectively. Ponderosa pine (Pinus ponderosa) needles contained higher values of {sup 238}Pu, {sup 239}Pu, and total uranium than did leaves collected from gambel`s oak (Quercus gambelii). In contrast, leaves collected from gambel`s oak contained higher levels of {sup 137}Cs than what pine needles did.

Fresquez, P.R.; Ennis, M.

1995-09-01T23:59:59.000Z

34

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

SciTech Connect (OSTI)

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.

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

1986-01-01T23:59:59.000Z

35

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

E-Print Network [OSTI]

,4]. The tritium is removed from the heavy water moderator at the Darlington Tritium Removal Facility (DTRF). for the availabil- ity of tritium to supply the requirements for the DT physics devices and power plants is closing

Abdou, Mohamed

36

Plutonium-Uranium Extraction (PUREX) facility preclosure work plan  

SciTech Connect (OSTI)

The dangerous waste permit identification number (WA7890008967)was issued by the U.S. Environmental Protection Agency and the Washington State Department of Ecology. This identification number encompasses a number of treatment, storage, and/or disposal units within the Hanford Facility. One of these treatment, storage, and/or disposal units is the PUREX Facility,currently undergoing a phased closure. The PUREX Facility Preclosure Work Plan submittal differs from closure plans previously submitted by the U.S. Department of Energy, Richland Operations Office to the Washington State Department of Ecology,in that the closure process occurs in three distinct phases as part of the decommissioning process (i.e., transition,surveillance and maintenance, and disposition). Final closure will occur during the disposition phase. This phased decommissioning process is implemented because development of a complete closure plan during the transition phase is impractical and future land use determinations have not been identified. The objective of the transition phase is to place the PUREX Facility in a safe configuration with respect to human health and the environment. Following the transition phase activities, the PUREX Facility will begin the surveillance and maintenance phase of 10 or more years until disposition phase activities commence. The closure plan for the PUREX facility will be prepared during the disposition phase. For purposes of this documentation, the PUREX Facility does not include the PUREX Storage Tunnels. The PUREX Storage Tunnels are an operating storage unit(DOE/RL-94-24).

Bhatia, R.K., Westinghouse Hanford

1996-07-09T23:59:59.000Z

37

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

SciTech Connect (OSTI)

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.

C.P.C. Wong; B. Merrill

2014-10-01T23:59:59.000Z

38

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]

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

2006-01-01T23:59:59.000Z

39

Facility effluent monitoring plan for the plutonium-uranium extraction facility  

SciTech Connect (OSTI)

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.

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

40

Behavior of tritium in the Purex process  

SciTech Connect (OSTI)

The tri-n-butyl phosphate(TBP) extraction behavior of tritium was studied to support the evaluation of tritium confinement in the Purex process. The tritium distribution ratio in the system of U - 30% TBP/n-dodecane(nDD) - HNO[sub 3] [minus]H[sub 2]O was measured in batch-wise experiments. The tritium decontamination factor in a tritium scrubbing step was measured in chemical flow sheet experiments using a 6 stage mixer-settler having an internal recycle system of aqueous solution in each stage. Tritium was mainly extracted in the forms of water and nitric acid in the system. Less than 1% of the amount of tritium in the organic phase was fixed in the degradation products of the solvent. A tritium decontamination factor of about 500 in the tritium scrubbing step was obtained under the conditions of about 85 g-U/dm[sup 3] of uranium concentration in the irradiated solvent(30% TBP/nDD), 3 M nitric acid in the tritium scrubbing solution and an organic: aqueous phase ratio of 25. 17 refs., 14 figs., 1 tab.

Uchiyama, Gunzo; Fujine, Sachio; Maeda, Mitsuru; Sugikawa, Susumu; Tsujino, Takeshi (Japan Atomic Eenrgy Inst., Ibaraki (Japan))

1995-01-01T23:59:59.000Z

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


41

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

42

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

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

November 13, 2013 Review of Savannah River Field Office Tritium Facilities Radiological Controls Activity-Level Implementation This report documents the results of an independent...

43

Universal tritium transmitter  

SciTech Connect (OSTI)

At the Savannah River Site and throughout the National Nuclear Security Agency (NNSA) tritium is measured using Ion or Kanne Chambers. Tritium flowing through an Ion Chamber emits beta particles generating current flow proportional to tritium radioactivity. Currents in the 1 x 10{sup -15} A to 1 x 10{sup -6} A are measured. The distance between the Ion Chamber and the electrometer in NNSA facilities can be over 100 feet. Currents greater than a few micro-amperes can be measured with a simple modification. Typical operating voltages of 500 to 1000 Volts and piping designs require that the Ion Chamber be connected to earth ground. This grounding combined with long cable lengths and low currents requires a very specialized preamplifier circuit. In addition, the electrometer must be able to supply 'fail safe' alarm signals which are used to alert personnel of a tritium leak, trigger divert systems preventing tritium releases to the environment and monitor stack emissions as required by the United States federal Government and state governments. Ideally the electrometer would be 'self monitoring'. Self monitoring would reduce the need for constant checks by maintenance personnel. For example at some DOE facilities monthly calibration and alarm checks must be performed to ensure operation. NNSA presently uses commercially available electrometers designed specifically for this critical application. The problems with these commercial units include: ground loops, high background currents, inflexibility and susceptibility to Electromagnetic Interference (EMI) which includes RF and Magnetic fields. Existing commercial electrometers lack the flexibility to accommodate different Ion Chamber designs required by the gas pressure, type of gas and range. Ideally the electrometer could be programmed for any expected gas, range and high voltage output. Commercially available units do not have 'fail safe' self monitoring capability. Electronics used to measure extremely low current must have sufficient time to thermally equilibrate. Amplifiers, transistors, resistors all need time to stabilize before the electrometer circuit will measure accurately in the 10{sup -15} and 10{sup -14} ampere range. Existing electrometers give the user no indication when the unit has stabilized and is acceptable for low level measurements. Savannah River National Laboratory (SRNL) funded through the NNSA Plant Directed Research and Development (PDRD) program, has developed a truly Universal Tritium Transmitter (UTT) capable of solving many known problems with existing commercial electrometers. This UTT pushes the state-of-the-art in electrometer design and incorporates solutions to deficiencies found in commercial electrometers. (authors)

Cordaro, J. V.; Wood, M. [Westinghouse Savannah River Company, Savannah River National Laboratory, Aiken, SC 29808 (United States)

2008-07-15T23:59:59.000Z

44

The control of tritium in ETHEL  

SciTech Connect (OSTI)

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.

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

1994-03-01T23:59:59.000Z

45

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

SciTech Connect (OSTI)

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.

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

1997-07-24T23:59:59.000Z

46

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...

47

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

SciTech Connect (OSTI)

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)

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

48

Tritium Management Control of tritium inventory is  

E-Print Network [OSTI]

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

Princeton Plasma Physics Laboratory

49

Continuous aqueous tritium monitor  

DOE Patents [OSTI]

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.

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

1989-05-30T23:59:59.000Z

50

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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarlyEnergy 0611__Joint_DOE_GoJ_AMS_Data_v3.pptx More Documents &DOE.F 1325.8CHANGE NOTICE NO.Change

51

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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarlyEnergy 0611__Joint_DOE_GoJ_AMS_Data_v3.pptx More Documents &DOE.F 1325.8CHANGE NOTICE

52

E-Print Network 3.0 - atmospheric tritium gas Sample Search Results  

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

+ D2O The degree of tritium and protium extraction... of the steam and gas mixture, the heavy water flow from the third stage of the CIE unit depleted of tritium... for hydrogen...

53

REMOTE ANALYSIS OF HIGH-TRITIUM-CONTENT WATER  

SciTech Connect (OSTI)

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.

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

2007-08-07T23:59:59.000Z

54

Continuous aqueous tritium monitor  

DOE Patents [OSTI]

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.

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

1987-10-19T23:59:59.000Z

55

TRITIUM SYSTEMS KEYWORDS: tritium fuel cycle, re-  

E-Print Network [OSTI]

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

Abdou, Mohamed

56

PDRD (SR13046) TRITIUM PRODUCTION FINAL REPORT  

SciTech Connect (OSTI)

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.

Smith, P.; Sheetz, S.

2013-09-30T23:59:59.000Z

57

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...

58

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

SciTech Connect (OSTI)

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)

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

59

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)

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.

NONE

1996-04-01T23:59:59.000Z

60

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

SciTech Connect (OSTI)

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.

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

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


61

Tritium transport in the NuMI decay pipe region - modeling and comparison with experimental data  

SciTech Connect (OSTI)

The NuMI (Neutrinos at Main Injector) beam facility at Fermilab is designed to produce an intense beam of muon neutrinos to be sent to the MINOS underground experiment in Soudan, Minnesota. Neutrinos are created by the decay of heavier particles. In the case of NuMI, the decaying particles are created by interaction of high-energy protons in a target, creating mostly positive pions. These particles can also interact with their environment, resulting in production of a variety of short-lived radionuclides and tritium. In the NuMI beam, neutrinos are produced by 120 GeV protons from the Fermilab Main Injector accelerator which are injected into the NuMI beam line using single turn extraction. The beam line has been designed for 400 kW beam power, roughly a factor of 2 above the initial (2005-06) running conditions. Extracted protons are bent downwards at a 57mr angle towards the Soudan Laboratory. The meson production target is a 94 cm segmented graphite rod, cooled by water in stainless tubes on the top and bottom of the target. The target is followed by two magnetic horns which are pulsed to 200 kA in synchronization with the passage of the beam, producing focusing of the secondary hadron beam and its daughter neutrinos. Downstream of the second horn the meson beam is transported for 675 m in an evacuated 2 m diameter beam (''decay'') pipe. Subsequently, the residual mesons and protons are absorbed in a water cooled aluminum/steel absorber immediately downstream of the decay pipe. Some 200 m of rock further downstream ranges out all of the residual muons. During beam operations, after installation of the chiller condensate system in December 2005, the concentration of tritiated water in the MINOS sump flow of 177 gpm was around 12 pCi/ml, for a total of 0.010 pCi/day. A simple model of tritium transport and deposition via humidity has been constructed to aid in understanding how tritium reaches the sump water. The model deals with tritium transported as HTO, water in which one hydrogen atom has been replaced with tritium. Based on concepts supported by the modeling, a dehumidification system was installed during May 2006 that reduced the tritium level in the sump by a factor of two. This note is primarily concerned with tritium that was produced in the NuMI target pile, carried by air flow into the target hall and down the decay pipe passageway (where most of it was deposited). The air is exhausted through the existing air vent shaft EAV2 (Figure 1).

Hylen, J.; Plunkett, R.; /Fermilab

2007-03-01T23:59:59.000Z

62

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)

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.

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

1998-11-10T23:59:59.000Z

63

Methods for tritium labeling  

DOE Patents [OSTI]

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.

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

1993-01-01T23:59:59.000Z

64

Oxidative Tritium Decontamination System  

DOE Patents [OSTI]

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.

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

2006-02-07T23:59:59.000Z

65

FINAL REPORT FOR TRITIUM WATER MONITOR  

SciTech Connect (OSTI)

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.

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

2011-04-25T23:59:59.000Z

66

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

67

Tritium breeding blanket  

SciTech Connect (OSTI)

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.

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

68

Tritium waste package  

DOE Patents [OSTI]

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.

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

69

Tritium waste package  

DOE Patents [OSTI]

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.

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

1995-11-07T23:59:59.000Z

70

2009 EVALUATION OF TRITIUM REMOVAL AND MITIGATION TECHNOLOGIES FOR WASTEWATER TREATMENT  

SciTech Connect (OSTI)

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.

LUECK KJ; GENESSE DJ; STEGEN GE

2009-02-26T23:59:59.000Z

71

MODELING ATMOSPHERIC RELEASES OF TRITIUM FROM NUCLEAR INSTALLATIONS  

SciTech Connect (OSTI)

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.

Okula, K

2007-01-17T23:59:59.000Z

72

Drum bubbler tritium processing system  

DOE Patents [OSTI]

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.

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

1999-01-01T23:59:59.000Z

73

Tritium deposition patterns in TFTR  

E-Print Network [OSTI]

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

Princeton Plasma Physics Laboratory

74

Monitoring of tritium  

DOE Patents [OSTI]

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.

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

1981-01-01T23:59:59.000Z

75

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

SciTech Connect (OSTI)

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.

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

2012-01-09T23:59:59.000Z

76

Drum bubbler tritium processing system  

DOE Patents [OSTI]

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.

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

1999-08-17T23:59:59.000Z

77

Tritium APEX Interim Report November, 1999  

E-Print Network [OSTI]

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

California at Los Angeles, University of

78

The certification process for tritium operators at TFTR  

SciTech Connect (OSTI)

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

Gentile, C.A.; Murphy, S.E.; LaMarche, P.H.; Contino, A.M.; Gordon, S. [Princeton Plasma Physics Lab., NJ (United States)

1995-12-31T23:59:59.000Z

79

Tritium monitor and collection system  

DOE Patents [OSTI]

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.

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

80

Tritium monitor and collection system  

DOE Patents [OSTI]

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.

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

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


81

Producing tritium in a homogenous reactor  

DOE Patents [OSTI]

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.

Cawley, William E. (Richland, WA)

1985-01-01T23:59:59.000Z

82

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

83

Operating Experience Review of Tritium-in-Water Monitors  

SciTech Connect (OSTI)

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.

S. A. Bruyere; L. C. Cadwallader

2011-09-01T23:59:59.000Z

84

TRITIUM ACCOUNTANCY IN FUSION SYSTEMS  

SciTech Connect (OSTI)

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.

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

85

Tritium-field betacells  

SciTech Connect (OSTI)

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.

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

86

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

SciTech Connect (OSTI)

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

Dominick, J L; Rasmussen, C L

2008-07-23T23:59:59.000Z

87

Composition containing aerogel substrate loaded with tritium  

DOE Patents [OSTI]

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.

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

88

Differential atmospheric tritium sampler  

DOE Patents [OSTI]

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.

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

1987-10-02T23:59:59.000Z

89

Differential atmospheric tritium sampler  

DOE Patents [OSTI]

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.

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

1990-01-01T23:59:59.000Z

90

Mercury and tritium removal from DOE waste oils  

SciTech Connect (OSTI)

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

Klasson, E.T. [Oak Ridge National Lab., TN (United States)

1997-10-01T23:59:59.000Z

91

Individual and population dose to users of the Savannah River following K-Reactor tritium release  

SciTech Connect (OSTI)

Approximately 5700 curies of tritium were released to Pen Branch between December 22, 1991 and December 25, 1991. As expected, the tritiated water traveled through the Savannah River swamp to Steel Creek and the Savannah River. Elevated tritium concentrations in the river at Becks Ferry (Beaufort-Jasper) and Abercorn Creek (Port Wentworth) has caused some concern among downstream water users as to the amount of tritium available for uptake through the domestic drinking water supplies at the Beaufort-Jasper and Port Wentworth water treatment facilities. Radiation dose to the downstream drinking water population is estimated in this report.

Carlton, W.H.; Hamby, D.M.

1992-12-31T23:59:59.000Z

92

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

SciTech Connect (OSTI)

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.

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

2005-07-15T23:59:59.000Z

93

Primer on tritium safe handling practices  

SciTech Connect (OSTI)

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.

Not Available

1994-12-01T23:59:59.000Z

94

Laser-assisted isotope separation of tritium  

DOE Patents [OSTI]

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.

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

1983-01-01T23:59:59.000Z

95

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

96

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

E-Print Network [OSTI]

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

Aeschbach-Hertig, Werner

97

Overview of Recent Tritium Experiments in TPE  

SciTech Connect (OSTI)

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.

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

2010-10-01T23:59:59.000Z

98

Neutrino mass limit from tritium beta decay  

E-Print Network [OSTI]

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

E. W. Otten; C. Weinheimer

2009-09-11T23:59:59.000Z

99

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

SciTech Connect (OSTI)

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.

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

2012-02-29T23:59:59.000Z

100

Status and practicality of detritiation and tritium production strategies for environmental remediation  

SciTech Connect (OSTI)

Operation of nuclear facilities throughout the world generates wastewater, groundwater and surface water contaminated with tritium. Because of a commitment to minimize radiation exposures to ''levels as low as reasonably achievable'', the US Department of Energy supports development of tritium isotope separation technologies. Also, DOE periodically documents the status and potential viability of alternative tritium treatment technologies and management strategies. The specific objectives of the current effort are to evaluate practical engineering issues, technology acceptability issues, and costs for realistic tritium treatment scenarios. A unique feature of the assessment is that the portfolio of options was expanded to include various management strategies rather than only evaluating detritiation technologies. The ultimate purpose of this effort is to assist Environmental Restoration and its support organizations in allocating future investments.

Fulbright, H.H.; Schwirian-Spann, A.L.; Brunt, V. van [Univ. of South Carolina, Columbia, SC (US); Jerome, K.M.; Looney, B.B. [Westinghouse Savannah River Co., Aiken, SC (US)

1996-02-26T23:59:59.000Z

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


101

Tritium hazard via the ingestion pathway  

SciTech Connect (OSTI)

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.

Travis, C.C.

1985-01-01T23:59:59.000Z

102

EFFECTS OF TRITIUM GAS EXPOSURE ON POLYMERS  

SciTech Connect (OSTI)

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.

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

2011-01-07T23:59:59.000Z

103

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

E-Print Network [OSTI]

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...

Bodine, L I; Robertson, R G H

2015-01-01T23:59:59.000Z

104

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

SciTech Connect (OSTI)

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.

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

2005-07-15T23:59:59.000Z

105

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;

106

DEPLOYMENT OF THE BULK TRITIUM SHIPPING PACKAGE  

SciTech Connect (OSTI)

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.

Blanton, P.

2013-10-10T23:59:59.000Z

107

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)

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.

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

2006-10-25T23:59:59.000Z

108

Tritium radioluminescent devices, Health and Safety Manual  

SciTech Connect (OSTI)

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.

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

1995-06-01T23:59:59.000Z

109

Recovery of tritium from tritiated molecules  

DOE Patents [OSTI]

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.

Swansiger, W.A.

1984-10-17T23:59:59.000Z

110

TRITIUM INCORPORATION STUDIES IN PHOTO-SYNTHETIC BACTERIA  

E-Print Network [OSTI]

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

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

2008-01-01T23:59:59.000Z

111

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...

112

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...

113

Reclassification of the Tritium Research Laboratory  

SciTech Connect (OSTI)

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

Johnson, A.J.

1997-01-01T23:59:59.000Z

114

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

SciTech Connect (OSTI)

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.

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

115

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

SciTech Connect (OSTI)

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)

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

116

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)

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).

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

2001-08-24T23:59:59.000Z

117

TRITIUM EFFECTS ON WELDMENT FRACTURE TOUGHNESS  

SciTech Connect (OSTI)

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''.

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

2006-07-17T23:59:59.000Z

118

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)

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)

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

119

Portable Intelligent Tritium in Air Monitor  

SciTech Connect (OSTI)

The tritium detection method used for this monitor is original, patented in Romania. The detection unit consists of a single ionization chamber, a special fast preamplifier and a dedicated software associated to the detection unit, for signals processing. Some results concerning the tritium in relative strong gamma-ray fields are presented.

Purghel, L. [National Institute for Physics and Nuclear Engineering (Romania); Calin, M.R. [National Institute for Physics and Nuclear Engineering (Romania); Bartos, D. [National Institute for Physics and Nuclear Engineering (Romania); Serbia, L. [National Institute for Physics and Nuclear Engineering (Romania); Lupu, A. [S.C. TEHNOPLUS S.R.L. (Romania); Lupu, F. [S.C. TEHNOPLUS S.R.L. (Romania); Lupu, D. [S.C. TEHNOPLUS S.R.L. (Romania)

2005-07-15T23:59:59.000Z

120

Vanadium hydride deuterium-tritium generator  

DOE Patents [OSTI]

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.

Christensen, Leslie D. (Livermore, CA)

1982-01-01T23:59:59.000Z

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


121

HYLIFE-II tritium management system  

SciTech Connect (OSTI)

The tritium management system performs seven functions: (1) tritium gas removal from the blast chamber, (2) tritium removal from the Flibe, (3) tritium removal from helium sweep gas, (4) tritium removal from room air, (5) hydrogen isotope separation, (6) release of non-hazardous gases through the stack, (7) fixation and disposal of hazardous effluents. About 2 TBq/s (5 MCi/day) of tritium is bred in the Flibe (Li{sub 2}BeF{sub 4}) molten salt coolant by neutron absorption. Tritium removal is accomplished by a two-stage vacuum disengager in each of three steam generator loops. Each stage consists of a spray of 0.4 mm diameter, hot Flibe droplets into a vacuum chamber 4 m in diameter and 7 m tall. As droplets fall downward into the vacuum, most of the tritium diffuses out and is pumped away. A fraction {Phi}{approx}10{sup {minus}5} of the tritium remains in the Flibe as it leaves the second stage of the vacuum disengager, and about 24% of the remaining tritium penetrates through the steam generator tubes, per pass, so the net leakage into the steam system is about 4.7 MBq/s (11 Ci/day). The required Flibe pumping power for the vacuum disengager system is 6.6 MW. With Flibe primary coolant and a vacuum disengager, an intermediate coolant loop is not needed to prevent tritium from leaking into the steam system. An experiment is needed to demonstrate vacuum disengager operation with Flibe. A secondary containment shell with helium sweep gas captures the tritium permeating out of the Flibe ducts, limiting leaks there to about 1 Ci/day. The tritium inventory in the reactor is about 190 g, residing mostly in the large Flibe recirculation duct walls. The total cost of the tritium management system is 92 M$, of which the vacuum disengagers cost = 56%, the blast chamber vacuum system = 15%, the cryogenic plant = 9%, the emergency air cleanup and waste treatment systems each = 6%, the protium removal system = 3%, and the fuel storage system and inert gas system each = 2%.

Longhurst, G.R.; Dolan, T.J.

1993-06-01T23:59:59.000Z

122

DOE handbook: Tritium handling and safe storage  

SciTech Connect (OSTI)

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.

NONE

1999-03-01T23:59:59.000Z

123

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

SciTech Connect (OSTI)

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.

Brown, D.H.

1998-06-30T23:59:59.000Z

124

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

SciTech Connect (OSTI)

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.

Klein, J.; Estochen, E.

2014-03-06T23:59:59.000Z

125

Tritium monitor with improved gamma-ray discrimination  

DOE Patents [OSTI]

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.

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

1982-10-21T23:59:59.000Z

126

Tritium monitor with improved gamma-ray discrimination  

DOE Patents [OSTI]

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.

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

1985-01-01T23:59:59.000Z

127

Tritium laboratory with multiple purposes at NIPNE Magurele Romania  

SciTech Connect (OSTI)

The Tritium Laboratory from NIPNE (Romania)) is part of Radioisotope Research and Production Center. The Tritium Laboratory has been in operation since 1960, and carries out R and D activities involving tritium sources in gaseous, liquids and solid state, provides specialized service to CANDU NPP Cernavoda (Romania)), and provides tritium assay services to internal and external customers. The paper presents the activities and perspectives of Tritium Laboratory and its performances in accordance with Quality System Management. (authors)

Matei, L.; Postolache, C. [Horia Hulubei, National Inst. for Physics and Nuclear Engineering NIPNE, 407 Atomistilor street, 077125 Magurele Ilfov (Romania)

2008-07-15T23:59:59.000Z

128

Tritium containing polymers having a polymer backbone substantially void of tritium  

DOE Patents [OSTI]

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.

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

1992-01-01T23:59:59.000Z

129

Tritium containing polymers having a polymer backbone substantially void of tritium  

DOE Patents [OSTI]

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.

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

1992-03-31T23:59:59.000Z

130

Overview of tritium processing development at the tritium systems test assembly  

SciTech Connect (OSTI)

The Tritium Systems Test Assembly (TSTA) at the Los Alamos National Laboratory has been operating with tritium since June 1984. Presently there are some 50 g of tritium in the main processing loop. This 50 g has been sufficient to do a number of experiments involving the cryogenic distillation isotope separation system and to integrate the fuel cleanup system into the main fuel processing loop. In January 1986 two major experiments were conducted. During these experiments the fuel cleanup system was integrated, through the transfer pumping system, with the isotope separation system, thus permitting testing on the integrated fuel processing loop. This integration of these systems leaves only the main vacuum system to be integrated into the TSTA fuel processing loop. In September 1986 another major tritium experiment was performed in which the integrated loop was operated, the tritium inventory increased to 50 g and additional measurements on the performance of the distillation system were taken. In the period June 1984 through September 1986 the TSTA system has processed well over 10/sup 8/ Ci of tritium. Total tritium emissions to the environment over this period have been less than 15 Ci. Personnel exposures during this period have totaled less than 100 person-mRem. To date, the development of tritium technology at TSTA has proceeded in progressive and orderly steps. In two years of operation with tritium, no major design flows have been uncovered.

Anderson, J.L.

1986-10-22T23:59:59.000Z

131

Tritium proof-of-principle injector experiment  

SciTech Connect (OSTI)

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.

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

1988-01-01T23:59:59.000Z

132

Development of Tritium Permeation Analysis Code (TPAC)  

SciTech Connect (OSTI)

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.

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

2010-10-01T23:59:59.000Z

133

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...

134

Vanadium hydride deuterium-tritium generator  

DOE Patents [OSTI]

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.

Christensen, L.D.

1980-03-13T23:59:59.000Z

135

Tritium level along Romanian Black Sea Coast  

SciTech Connect (OSTI)

Establishing the tritium level along the Romanian Black Sea Coast, after 10 years of exploitation of the nuclear power plant from Cernavoda, is a first step in evaluating its impact on the Black Sea ecosystem. The monitoring program consists of tritium activity concentration measurement in sea water and precipitation from Black Sea Coast between April 2005 and April 2006. The sampling points were spread over the Danube-Black Sea Canal - before the locks Agigea and Navodari, and Black Sea along the coast to the Bulgarian border. The average tritium concentration in sea water collected from the sampling locations had the value of 11.1 {+-} 2.1 TU, close to tritium concentration in precipitation. Although an operating nuclear power plant exists in the monitored area, the values of tritium concentration in two locations are slightly higher than those recorded elsewhere. To conclude, it could be emphasized that until now, Cernavoda NPP did not had any influence on the tritium concentration of the Black Sea Shore. (authors)

Varlam, C.; Stefanescu, I.; Popescu, I.; Faurescu, I. [National Inst. for Cryogenic and Isotopic Technologies, PO Box 10, Rm. Valcea, 24050 (Romania)

2008-07-15T23:59:59.000Z

136

Independent Oversight Review, Savannah River Site 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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarly Career Scientists'Montana.ProgramJulietipDepartment ofTheDepartment of

137

Independent Oversight Review, Savannah River Site Tritium Facilities - June  

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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarly Career Scientists'Montana.ProgramJulietipDepartment ofTheDepartment of2012 | Department of Energy

138

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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary)morphinanInformation Desert Southwest RegionatSearchScheduled System BurstLongTitanConsortium UniversitiesSavannah

139

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

E-Print Network [OSTI]

& tritium. The nuclear environment also affectsIn-vessel Components The nuclear environment also affects with an extensive R&D Program toward multiple- effect and integrated testing in fusion facilities (ITER TBM and FNSF ­ Functions, environmental conditions, integration with nuclear components ­ History of Blanket design

Abdou, Mohamed

140

Computational modeling and analysis of airflow in a tritium storage room  

SciTech Connect (OSTI)

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.

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

2003-01-01T23:59:59.000Z

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


141

System support software for TSTA (Tritium Systems Test Assembly)  

SciTech Connect (OSTI)

The fact that Tritium Systems Test Assembly (TSTA) is an experimental facility makes it impossible and undesirable to try to forecast the exact software requirements. Thus the software had to be written in a manner that would allow modifications without compromising the safety requirements imposed by the handling of tritium. This suggested a multi-level approach to the software. In this approach (much like the ISO network model) each level is isolated from the level below and above by cleanly defined interfaces. For example, the subsystem support level interfaces with the subsystem hardware through the software support level. Routines in the software support level provide operations like ''OPEN VALVE'' and CLOSE VALVE'' to the subsystem level. This isolates the subsystem level from the actual hardware. This is advantageous because changes can occur in any level without the need for propagating the change to any other level. The TSTA control system consists of the hardware level, the data conversion level, the operator interface level, and the subsystem process level. These levels are described.

Claborn, G.W.; Mann, L.W.; Nielson, C.W.

1987-10-01T23:59:59.000Z

142

Method and apparatus for controlling accidental releases of tritium  

DOE Patents [OSTI]

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.

Galloway, T.R.

1980-04-01T23:59:59.000Z

143

Recovery of tritium from a liquid lithium blanket  

SciTech Connect (OSTI)

The sorption of tritium on yttrium from liquid lithium and the subsequent release of tritium from yttrium by thermal regeneration of the metal sorbent were investigated to study such a tritium-recovery process for a fusion reactor blanket of liquid lithium. Recent static sorption experiments have shown the effects of lithium temperature and possible impurities on the sorption of tritium. Diffusivity data, obtained from previous tritium recovery experiments, were evaluated to show the importance of the yttrium surface condition in controlling the release of tritium.

Talbot, J.B.

1981-01-01T23:59:59.000Z

144

Method and apparatus for controlling accidental releases of tritium  

DOE Patents [OSTI]

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.

Galloway, Terry R. [Berkeley, CA

1980-04-01T23:59:59.000Z

145

A compact tritium AMS system  

SciTech Connect (OSTI)

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}.

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

1999-09-23T23:59:59.000Z

146

Former Reactor Facilities Surveillance and Maintenance and  

E-Print Network [OSTI]

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

Ohta, Shigemi

147

Low technology high tritium breeding blanket concept  

SciTech Connect (OSTI)

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.

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

148

Selection of fluids for tritium pumping systems  

SciTech Connect (OSTI)

The degradation characteristics of three types of vacuum pump fluids, polyphenyl ethers, perfluoropolyethers and hydrocarbon oils were reviewed. Fluid selection proved to be a critical factor in the long-term performance of tritium pumping systems and subsequent tritium recovery operations. Thermal degradation and tritium radiolysis of pump fluids produce contaminants which can damage equipment and interfere with tritium recovery operations. General characteristics of these fluids are as follows: polyphenyl ether has outstanding radiation resistance, is very stable under normal diffusion pump conditions, but breaks down in the presence of oxygen at anticipated operating temperatures. Perfluoropolyether fluids are very stable and do not react chemically with most gases. Thermal and mechanical degradation products are inert, but the radiolysis products are very corrosive. Most of the degradation products of hydrogen oils are volatile and the principal radiolysis product is methane. Our studies show that polyphenyl ethers and hydrocarbon oils are the preferred fluids for use in tritium pumping systems. No corrosive materials are formed and most of the degradation products can be removed with suitable filter systems.

Chastagner, P

1984-02-01T23:59:59.000Z

149

Analysis of tritium transport in irradiated beryllium  

SciTech Connect (OSTI)

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.

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

1994-12-31T23:59:59.000Z

150

Results of Tritium Tracking and Groundwater Monitoring at the Hanford Site 200 Area State-Approved Land Disposal Site-FY1999  

SciTech Connect (OSTI)

The Hanford Site 200 Area Effluent Treatment Facility (ETF) processes contaminated liquids derived from Hanford Site facilities. The clean water generated by these processes is occasionally enriched in tritium and is discharged to the 200 Area State Approved Land Disposal Site (SALDS). Groundwater monitoring for tritium and other constituents is required by the state-issued permit at 21 wells surrounding the facility. During FY 1999, average tritium activities in most wells declined from average activities in 1998. The exception was deep well 69948-77C, where tritium results were at an all-time high (77,000 pCi/L) as a result of the delayed penetration of effluent deeper into the aquifer. Of the 12 constituents with permit enforcement limits, which are monitored in SALDS proximal wells, all were within limits during FY 1999. Water level measurements in nearby wells indicate that a small hydraulic mound exists around the SALDS facility as a result of discharges. This feature is directing groundwater flow radially outward a short distance before the regional northeasterly flow predominates. Evaluation of this condition indicates that the network is currently adequate for tracking potential effects of the SALDS on the groundwater. Recommendations include the discontinuation of ammonia, benzene, tetrahydrofuran, and acetone from the regular groundwater constituent list; designating background well 299-W8-1 as a tritium-tracking well only, and the use of quadruplicate averages of field pH, instead of a single laboratory measurement, as a permit compliance parameter.

Barnett, D.B.

1999-10-20T23:59:59.000Z

151

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...

152

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

E-Print Network [OSTI]

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.

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

2015-02-12T23:59:59.000Z

153

Update: tritium at Fermilab Fermilab Community Advisory Board  

E-Print Network [OSTI]

, 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. · Becomes part of water molecules like normal hydrogen · Cannot penetrate skin. · Does not accumulate

Quigg, Chris

154

Separation phenomenon in the Windowless Gaseous Tritium Source of KATRIN  

E-Print Network [OSTI]

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

Sharipov, Felix

155

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

E-Print Network [OSTI]

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

Aeschbach-Hertig, Werner

156

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.

157

Small system for tritium accelerator mass spectrometry  

DOE Patents [OSTI]

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.

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

1993-01-01T23:59:59.000Z

158

Small system for tritium accelerator mass spectrometry  

DOE Patents [OSTI]

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.

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

1993-02-23T23:59:59.000Z

159

Tritium management in fusion synfuel designs  

SciTech Connect (OSTI)

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.

Galloway, T.R.

1980-04-25T23:59:59.000Z

160

Tritium at Fermilab Fermilab Community Advisory Board  

E-Print Network [OSTI]

Tritium at Fermilab Fermilab Community Advisory Board September 23, 2010 Rob Plunkett, Fermilab #12;2 Got water? Robert Plunkett #12;Fermilab has plenty Robert Plunkett3 The Fermilab site has numerous ponds and is the origin of Indian Creek and Ferry Creek. Fermilab uses water to cool accelerators

Quigg, Chris

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


161

Tokamak fusion reactors with less than full tritium breeding  

SciTech Connect (OSTI)

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.

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

1983-05-01T23:59:59.000Z

162

DEVELOPMENT OF THE BULK TRITIUM SHIPPING PACKAGING  

SciTech Connect (OSTI)

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.

Blanton, P.; Eberl, K.

2008-09-14T23:59:59.000Z

163

IN-SITU TRITIUM BETA DETECTOR  

SciTech Connect (OSTI)

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.

J.W. Berthold; L.A. Jeffers

1998-04-15T23:59:59.000Z

164

Apparatus to recover tritium from tritiated molecules  

DOE Patents [OSTI]

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.

Swansiger, William A. (Livermore, CA)

1988-01-01T23:59:59.000Z

165

Facility effluent monitoring plan for the 324 Facility  

SciTech Connect (OSTI)

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

NONE

1994-11-01T23:59:59.000Z

166

Progress in tritium retention and release modeling for ceramic breeders  

SciTech Connect (OSTI)

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.

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

1994-12-31T23:59:59.000Z

167

Recovery of tritium from tritiated molecules  

DOE Patents [OSTI]

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.

Swansiger, William A. (Livermore, CA)

1987-01-01T23:59:59.000Z

168

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

E-Print Network [OSTI]

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.

Jordan, Preston D.; Javandel, Iraj

2007-01-01T23:59:59.000Z

169

Tritium inventory control in ITER Charles Skinner with key contributions from  

E-Print Network [OSTI]

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

Princeton Plasma Physics Laboratory

170

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

SciTech Connect (OSTI)

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.

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

171

TRITIUM BARRIER MATERIALS AND SEPARATION SYSTEMS FOR THE NGNP  

SciTech Connect (OSTI)

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.

Sherman, S; Thad Adams, T

2008-07-17T23:59:59.000Z

172

2012 ACCOMPLISHMENTS - TRITIUM AGING STUDIES ON STAINLESS STEELS  

SciTech Connect (OSTI)

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.

Morgan, M.

2013-01-31T23:59:59.000Z

173

EFFECTS OF TRITIUM GAS EXPOSURE ON EPDM ELASTOMER  

SciTech Connect (OSTI)

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.

Clark, E.

2009-12-11T23:59:59.000Z

174

Tritium Formation and Mitigation in High-Temperature Reactors  

SciTech Connect (OSTI)

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

Piyush Sabharwall; Carl Stoots

2012-10-01T23:59:59.000Z

175

Tritium Formation and Mitigation in High-Temperature Reactor Systems  

SciTech Connect (OSTI)

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

Piyush Sabharwall; Carl Stoots; Hans A. Schmutz

2013-03-01T23:59:59.000Z

176

E-Print Network 3.0 - adrenal hyperplasia tritium Sample Search...  

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

liquid scintillation analysis of tritium... which has been doped electrolytically in the bulk specimen. The tritium release showed quantitatively Source: Ecole Polytechnique,...

177

E-Print Network 3.0 - assess tritium levels Sample Search Results  

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...

178

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...

179

Mixed Waste Management Facility Groundwater Monitoring Report  

SciTech Connect (OSTI)

During fourth quarter 1997, eleven constituents exceeded final Primary Drinking Water Standards (PDWS) in groundwater samples from downgradient monitoring wells at the Mixed Waste Management Facility. No constituents exceeded final PDWS in samples from upgradient monitoring wells. As in previous quarters, tritium and trichloroethylene were the most widespread elevated constituents. The groundwater flow directions and rates in the three hydrostratigraphic units were similar to those of previous quarters.

Chase, J.

1998-03-01T23:59:59.000Z

180

Improving cryogenic deuterium–tritium implosion performance on OMEGA  

SciTech Connect (OSTI)

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.

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

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


181

Determination of tritium distribution in labeled compounds using EPR spectrometry  

SciTech Connect (OSTI)

Usually, the tritium distribution in a labeled compound is analyzed by T-NMR spectrometry. NMR equipment is expensive and its sensitivity is lower in comparison to EPR spectrometry. In this paper, the possibility of determining the distribution of tritium in a labeled molecule using self-radiolytic decay processes was analyzed. (authors)

Postolache, C.; Matei, L.; Georgescu, R. [Horia Hulubei, National Inst. for Physics and Nuclear Engineering IFIN HH, 407 Atomistilor street, 077125 Magurele Ilfov (Romania)

2008-07-15T23:59:59.000Z

182

Effectiveness Monitoring Report, MWMF Tritium Phytoremediation Interim Measures.  

SciTech Connect (OSTI)

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

Hitchcock, Dan; Blake, John, I.

2003-02-10T23:59:59.000Z

183

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

SciTech Connect (OSTI)

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.

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

1995-12-31T23:59:59.000Z

184

Progress report on the accelerator production of tritium materials irradiation program  

SciTech Connect (OSTI)

The Accelerator Production of Tritium (APT) project is developing an accelerator and a spoliation neutron source capable of producing tritium through neutron capture on He-3. A high atomic weight target is used to produce neutrons that are then multiplied and moderated in a blanket prior to capture. Materials used in the target and blanket region of an APT facility will be subjected to several different and mixed particle radiation environments; high energy protons (1-2 GeV), protons in the 20 MeV range, high energy neutrons, and low energy neutrons, depending on position in the target and blanket. Flux levels exceed 10{sup 14}/cm{sup 2}s in some areas. The APT project is sponsoring an irradiation damage effects program that will generate the first data-base for materials exposed to high energy particles typical of spallation neutron sources. The program includes a number of candidate materials in small specimen and model component form and uses the Los Alamos Spallation Radiation Effects Facility (LASREF) at the 800 MeV, Los Alamos Neutron Science Center (LANSCE) accelerator.

Maloy, S.A.; Sommer, W.F.; Brown, R.D.; Roberts, J.E. [and others

1997-05-01T23:59:59.000Z

185

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

186

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

187

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

188

EFFECT OF TRITIUM AND DECAY HELIUM ON WELDMENT FRACTURE TOUGHNESS  

SciTech Connect (OSTI)

The fracture toughness data collected in this study are needed to assess the long-term effects of tritium and its decay product on tritium reservoirs. The results show that tritium and decay helium have negative effects on the fracture toughness properties of stainless steel and its weldments. The data and report from this study has been included in a material property database for use in tritium reservoir modeling efforts like the Technology Investment Program ''Lifecycle Engineering for Tritium Reservoirs''. A number of conclusions can be drawn from the data: (1) For unexposed Type 304L stainless steel, the fracture toughness of weldments was two to three times higher than the base metal toughness. (2) Tritium exposure lowered the fracture toughness properties of both base metals and weldments. This was characterized by lower J{sub Q} values and lower J-da curves. (3) Tritium-exposed-and-aged base metals and weldments had lower fracture toughness values than unexposed ones but still retained good toughness properties.

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

2006-09-26T23:59:59.000Z

189

MEASUREMENT OF TRITIUM DURING VOLOXIDATION OF ZIRCALOY-2 FUEL HULLS  

SciTech Connect (OSTI)

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.

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

2010-10-14T23:59:59.000Z

190

Facility Microgrids  

SciTech Connect (OSTI)

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.

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

2005-05-01T23:59:59.000Z

191

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

SciTech Connect (OSTI)

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.

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

1995-09-01T23:59:59.000Z

192

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

DOE Patents [OSTI]

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.

Anderson, David F. (Los Alamos, NM)

1984-01-01T23:59:59.000Z

193

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

DOE Patents [OSTI]

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.

Anderson, D.F.

1981-01-27T23:59:59.000Z

194

Tritium Formation and Mitigation in High Temperature Reactors  

SciTech Connect (OSTI)

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.

Piyush Sabharwall; Carl Stoots

2012-08-01T23:59:59.000Z

195

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

SciTech Connect (OSTI)

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.

Senor, David J.

2013-10-30T23:59:59.000Z

196

On-line tritium production monitor  

DOE Patents [OSTI]

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.

Mihalczo, J.T.

1993-11-23T23:59:59.000Z

197

On-line tritium production monitor  

DOE Patents [OSTI]

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.

Mihalczo, John T. (Oak Ridge, TN)

1993-01-01T23:59:59.000Z

198

Ignition of deuterium-tritium fuel targets  

DOE Patents [OSTI]

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.

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

1991-08-27T23:59:59.000Z

199

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

SciTech Connect (OSTI)

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.

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

1995-08-01T23:59:59.000Z

200

Plasma Fueling, Pumping, and Tritium Handling Considerations for FIRE  

SciTech Connect (OSTI)

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.

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

1999-11-13T23:59:59.000Z

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


201

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...

202

Long Term Tritium Trapping in TFTR and JET  

SciTech Connect (OSTI)

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.

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

203

Progress in tritium retention and release modeling for ceramic breeders  

SciTech Connect (OSTI)

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.

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

204

Facility Safety  

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

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

1996-10-24T23:59:59.000Z

205

Facility Safety  

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

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

1995-11-16T23:59:59.000Z

206

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

SciTech Connect (OSTI)

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.

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

207

International Facility Management Association Strategic Facility  

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

Facility Management Association Strategic Facility Planning: A WhIte PAPer Strategic Facility Planning: A White Paper on Strategic Facility Planning 2009 | International...

208

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

SciTech Connect (OSTI)

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.

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

2011-02-23T23:59:59.000Z

209

Tritium plasma experiment: Parameters and potentials for fusion plasma-wall interaction studies  

SciTech Connect (OSTI)

The tritium plasma experiment (TPE) is a unique facility devoted to experiments on the behavior of deuterium/tritium in toxic (e.g., beryllium) and radioactive materials for fusion plasma-wall interaction studies. A Langmuir probe was added to the system to characterize the plasma conditions in TPE. With this new diagnostic, we found the achievable electron temperature ranged from 5.0 to 10.0 eV, the electron density varied from 5.0 x 10{sup 16} to 2.5 x 10{sup 18} m{sup -3}, and the ion flux density varied between 5.0 x 10{sup 20} to 2.5 x 10{sup 22} m{sup -2} s{sup -1} along the centerline of the plasma. A comparison of these plasma parameters with the conditions expected for the plasma facing components (PFCs) in ITER shows that TPE is capable of achieving most ({approx}800 m{sup 2} of 850 m{sup 2} total PFCs area) of the expected ion flux density and electron density conditions.

Shimada, Masashi; Sharpe, J. Phillip [Fusion Safety Program, Idaho National Laboratory, Idaho Falls, Idaho 83415 (United States); Kolasinski, Robert D.; Causey, Rion A. [Hydrogen and Metallurgical Science Department, Sandia National Laboratories, Livermore, California 94551 (United States)

2011-08-15T23:59:59.000Z

210

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

SciTech Connect (OSTI)

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

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

1994-06-01T23:59:59.000Z

211

Experimental Investigation in Order to Determine Catalytic Package Performances in Case of Tritium Transfer from Water to Gas  

SciTech Connect (OSTI)

The processes for hydrogen isotope's separation are very important for nuclear technology. One of the most important processes for tritium separation, is the catalyst isotope exchange water-hydrogen.Our catalytic package consists of Romanian patented catalysts with platinum on charcoal and polytetrafluoretylene (Pt/C/PTFE) and the ordered Romanian patented package B7 type. The catalytic package was tested in an isotope exchange facility for water detritiation at the Experimental Pilot Plant from ICIT Rm.Valcea.In a column of isotope exchange tritium is transferred from liquid phase (tritiated heavy water) in gaseous phase (hydrogen). In the experimental set-up, which was used, the column of catalytic isotope exchange is filled with successive layers of catalyst and ordered package. The catalyst consists of 95.5 wt.% of PTFE, 4.1 wt. % of carbon and 0.40 wt. % of platinum and was of Raschig rings 10 x 10 x 2 mm. The ordered package was B7 type consists of wire mesh phosphor bronze 4 x 1 wire and the mesh dimension is 0.18 x 0.48 mm.We analyzed the transfer phenomena of tritium from liquid to gaseous phase, in this system.The mass transfer coefficient which characterized the isotopic exchange on the package, were determined as function of experimental parameters.

Bornea, Anisia [National Institute of R and D for Cryogenics and Isotopic Separations (Romania); Peculea, M. [Romanian Academy (Romania); Zamfirache, M. [National Institute of R and D for Cryogenics and Isotopic Separations (Romania); Varlam, Carmen [National Institute of R and D for Cryogenics and Isotopic Separations (Romania)

2005-07-15T23:59:59.000Z

212

Tritium Effects on Fracture Toughness of Stainless Steel Weldments  

SciTech Connect (OSTI)

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.

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

2005-05-12T23:59:59.000Z

213

RADIOLYTIC GAS PRODUCTION RATES OF POLYMERS EXPOSED TO TRITIUM GAS  

SciTech Connect (OSTI)

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

Clark, E.

2013-08-31T23:59:59.000Z

214

E-Print Network 3.0 - aqueous effluent tritium Sample Search...  

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

tritium and 14 C may... that a higher tritium was observed in paint from the wall of the heavy water room (500 Bqg or 20 Bqcm2... (0.8 Bqg). During the operation of reactor,...

215

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

E-Print Network [OSTI]

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

Sharipov, Felix

216

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

DOE Patents [OSTI]

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.

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

1982-08-19T23:59:59.000Z

217

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...

218

Modeling Tritium Transport in PbLi Breeder Blankets Under Steady State  

SciTech Connect (OSTI)

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.

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

2011-08-01T23:59:59.000Z

219

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)

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.

Winterberg, F.

2009-10-29T23:59:59.000Z

220

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

E-Print Network [OSTI]

24 IV SYSTEMS STUDIES System Descriptions Parametric Studies Reference Case 25 25 27 30 RESULTS Fusile Inventory Hybrid Capital Costs System Sensitivity to Q Capital Costs for Tritium Production Reactors Doubling Time Effects Tritium... PAGE 26 28 III IV V System Sensitivity to the Fusile Inventory The Effect of the Tritium Production Reactor Capital Cost on the Cost of Electricity for the System The Effects on the System of Tritium Production in the Fission Power Reactors...

Rice, Brent Lee

1987-01-01T23:59:59.000Z

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


221

Dynamic simulation of a proposed ITER tritium processing system  

SciTech Connect (OSTI)

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.

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

222

Tritium Permeability of Incoloy 800H and Inconel 617  

SciTech Connect (OSTI)

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

Philip Winston; Pattrick Calderoni; Paul Humrickhouse

2011-09-01T23:59:59.000Z

223

Tritium Permeability of Incoloy 800H and Inconel 617  

SciTech Connect (OSTI)

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

Philip Winston; Pattrick Calderoni; Paul Humrickhouse

2012-07-01T23:59:59.000Z

224

Facility Safety  

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

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.

2005-12-22T23:59:59.000Z

225

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

226

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

E-Print Network [OSTI]

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

Abdou, Mohamed

227

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

E-Print Network [OSTI]

TRITIUM ANALYSIS OF A WATER-COOLED SOLID BREEDER BLANKET FOR ITER* G. Federici, A.R. Raffray, M breeder blanket for the InternationalThermonuclearExperimental Reac- tor (ITER) obtained from the tritium-cooled solid breeder blanket has been proposed for the ITER tritium-producing blanket [2]. The breeder is oper

Abdou, Mohamed

228

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

229

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

E-Print Network [OSTI]

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

Sharipov, Felix

230

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

231

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

232

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

233

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

E-Print Network [OSTI]

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

Slatton, Clint

234

DIFFUSION ELASTIQUE DES NEUTRONS PAR LE TRITIUM A 14 MeV  

E-Print Network [OSTI]

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

Paris-Sud XI, Université de

235

long-Term Tritium Transport through Field-Scale Compacted Soil Liner  

E-Print Network [OSTI]

long-Term Tritium Transport through Field-Scale Compacted Soil Liner Cecile Toupiol1; Thomas W. Daniel7 Abstract: A l3-year study of tritium transport through a field-scale earthen liner sampling) were used to determine the vertical tritium concentration profiles at different times

236

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

E-Print Network [OSTI]

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

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

237

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 injected into the vessel. ITER has focused considerable effort into the ability to thermally remove tritium vessel. Another possible technique to mitigate tritium accumulation in these codeposited surfaces

238

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

E-Print Network [OSTI]

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

Sharipov, Felix

239

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

E-Print Network [OSTI]

' . . , . 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

Harilal, S. S.

240

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

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


241

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

E-Print Network [OSTI]

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

Abdou, Mohamed

2006-01-01T23:59:59.000Z

242

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

243

Accelerator driven production of tritium: target and blanket design  

E-Print Network [OSTI]

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...

Ragusa, Jean Concetto

1996-01-01T23:59:59.000Z

244

Test of potential homogeneity in the KATRIN gaseous tritium source  

E-Print Network [OSTI]

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.

M. Rysavy

2005-06-02T23:59:59.000Z

245

Facility Safety  

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

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.

2002-05-20T23:59:59.000Z

246

Facility Safety  

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

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.

2013-06-21T23:59:59.000Z

247

Facility Safety  

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

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.

2005-12-22T23:59:59.000Z

248

Facility Safety  

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

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.

2000-11-20T23:59:59.000Z

249

Treatment of gaseous effluents at nuclear facilities  

SciTech Connect (OSTI)

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.

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

250

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

SciTech Connect (OSTI)

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)

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

2006-07-01T23:59:59.000Z

251

Mixed Waste Management Facility (MWMF) groundwater monitoring report. Fourth quarter 1992 and 1992 summary  

SciTech Connect (OSTI)

During fourth quarter 1992, nine constituents exceeded final Primary Drinking Water Standards (PDWS) in one or more groundwater samples from monitoring wells at the Mixed Waste Management Facility (MWMF) and adjacent facilities. As in previous quarters, tritium and trichloroethylene were the most widespread constituents. Fifty-seven (48%) of the 120 monitoring wells, contained elevated tritium activities, and 23 (19%) contained elevated trichloroethylene concentrations. Total alpha-emitting radium, tetrachloroethylene, chloroethene, cadmium, 1,1-dichloroethylene, lead, or nonvolatile beta levels exceeded standards in one or more wells. During 1992, elevated levels of 13 constituents were found in one or more of 80 of the 120 groundwater monitoring wells (67%) at the MWMF and adjacent facilities. Tritium and trichloroethylene exceeded their final PDWS more frequently and more consistently than did other constituents. Tritium activity exceeded its final PDWS m 67 wells and trichloroethylene was. elevated in 28 wells. Lead, tetrachloroethylene, total alpha-emitting radium, gross alpha, cadmium, chloroethene, 1,1-dichloroethylene 1,2-dichloroethane, mercury, or nitrate exceeded standards in one or more wells during the year. Nonvolatile beta exceeded its drinking water screening level in 3 wells during the year.

Not Available

1993-03-01T23:59:59.000Z

252

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

SciTech Connect (OSTI)

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.

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

253

Scoping Analyses on Tritium Permeation to VHTR Integarted Industrial Application Systems  

SciTech Connect (OSTI)

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

Chang H. Oh; Eung S. Kim

2011-03-01T23:59:59.000Z

254

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

E-Print Network [OSTI]

Science and Fusion Center, Cambridge, Massachusetts 02139, USA 2 Lawrence Livermore National Laboratory layered deuterium-tritium implosions on the NIF M. Gatu Johnson, D. T. Casey, J. A. Frenje, C.-K. Li, F. H layered deuterium-tritium implosions on the NIF M. Gatu Johnson,1,a) D. T. Casey,1,b) J. A. Frenje,1 C

255

Draft Supplemental Environmental Impact Statement for the Production...  

National Nuclear Security Administration (NNSA)

Facility at the Savannah River Site. The tritium is extracted in a high-temperature heating and vacuum process, after which it is purified. Impacts of Tritium Production on...

256

Facility Safety  

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

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.

2012-12-04T23:59:59.000Z

257

Facility Safety  

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

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.

1995-10-13T23:59:59.000Z

258

Silicon carbide tritium permeation barrier for steel structural components.  

SciTech Connect (OSTI)

Chemical vapor deposited (CVD) silicon carbide (SiC) has superior resistance to tritium permeation even after irradiation. Prior work has shown Ultrametfoam to be forgiving when bonded to substrates with large CTE differences. The technical objectives are: (1) Evaluate foams of vanadium, niobium and molybdenum metals and SiC for CTE mitigation between a dense SiC barrier and steel structure; (2) Thermostructural modeling of SiC TPB/Ultramet foam/ferritic steel architecture; (3) Evaluate deuterium permeation of chemical vapor deposited (CVD) SiC; (4) D testing involved construction of a new higher temperature (> 1000 C) permeation testing system and development of improved sealing techniques; (5) Fabricate prototype tube similar to that shown with dimensions of 7cm {theta} and 35cm long; and (6) Tritium and hermeticity testing of prototype tube.

Causey, Rion A. (Sandia National Laboratories, Livermore, CA); Garde, Joseph Maurico; Buchenauer, Dean A. (Sandia National Laboratories, Livermore, CA); Calderoni, Pattrick (Idaho National Laboratory); Holschuh, Thomas, Jr.; Youchison, Dennis Lee; Wright, Matt; Kolasinski, Robert D. (Sandia National Laboratories, Livermore, CA)

2010-09-01T23:59:59.000Z

259

Direct determination of Neutrino Mass from Tritium Beta Spectrum  

E-Print Network [OSTI]

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.

C. Weinheimer

2009-12-08T23:59:59.000Z

260

FINITE ELEMENT ANALYSIS OF BULK TRITIUM SHIPPING PACKAGE  

SciTech Connect (OSTI)

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.

Jordan, J.

2010-06-02T23:59:59.000Z

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


261

IN-LINE CHEMICAL SENSOR DEPLOYMENT IN A TRITIUM PLANT  

SciTech Connect (OSTI)

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.

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

2013-10-02T23:59:59.000Z

262

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

SciTech Connect (OSTI)

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

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

2013-09-01T23:59:59.000Z

263

Fast Flux Test Facility (FFTF) Briefing Book 1 Summary  

SciTech Connect (OSTI)

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

WJ Apley

1997-12-01T23:59:59.000Z

264

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

DOE Patents [OSTI]

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.

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

1989-12-26T23:59:59.000Z

265

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

DOE Patents [OSTI]

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.

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

1989-01-01T23:59:59.000Z

266

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

E-Print Network [OSTI]

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

Paris-Sud XI, Université de

267

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

SciTech Connect (OSTI)

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.

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

2005-08-30T23:59:59.000Z

268

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

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsrucLas Conchas recovery challengeMultiscaleLogos NERSCJeffreyKeyAdministration Breaks Ground on

269

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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarlyEnergy 0611__Joint_DOE_GoJ_AMS_Data_v3.pptx More Documents &DOE.F 1325.8CHANGE NOTICESuperseding

270

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

DOE Patents [OSTI]

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.

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

2001-01-01T23:59:59.000Z

271

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

E-Print Network [OSTI]

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... is readily assimilated by the body following an ingestion of HTO or by passage through the skin or I IOO/o P =ISkeV 2 He~ Figure 2. Radioactive Decay Scheme for Tritium 12 Table 1. Tritium Accumulation in 19B3 Atmospheric Weapons Testing* Natural...

Salsman, John Matthew

1983-01-01T23:59:59.000Z

272

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.  

273

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

SciTech Connect (OSTI)

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.

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

1993-12-31T23:59:59.000Z

274

CX-000547: Categorical Exclusion Determination | Department of...  

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

used to analyze particulate from inside two approximately 18 inch lengths of process piping from the Tritium Extraction Facility (TEF). DOCUMENT(S) AVAILABLE FOR DOWNLOAD...

275

--No Title--  

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

(TEM) will be used to analyze particulate from inside two 18" lengths of process piping from the Tritium Extraction Facility (TEF). Radiological survey results indicate the...

276

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...

277

Fluid extraction  

DOE Patents [OSTI]

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.

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

1999-01-01T23:59:59.000Z

278

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

SciTech Connect (OSTI)

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%.

Kuhne, W.

2012-12-03T23:59:59.000Z

279

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

DOE Patents [OSTI]

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.

Heung, Leung K. (Aiken, SC)

1994-01-01T23:59:59.000Z

280

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

DOE Patents [OSTI]

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.

Heung, L.K.

1994-03-29T23:59:59.000Z

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


281

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

SciTech Connect (OSTI)

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.

Ivey, Wade

2013-12-17T23:59:59.000Z

282

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

SciTech Connect (OSTI)

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

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

2014-02-15T23:59:59.000Z

283

EFFECTS OF ONE WEEK TRITIUM EXPOSURE ON EPDM ELASTOMER  

SciTech Connect (OSTI)

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.

Clark, E

2007-06-07T23:59:59.000Z

284

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

285

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

286

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

287

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

288

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

289

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

290

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

E-Print Network [OSTI]

.,, 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

Harilal, S. S.

291

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

E-Print Network [OSTI]

. & ,. 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

Harilal, S. S.

292

Lithium aluminate/zirconium material useful in the production of tritium  

DOE Patents [OSTI]

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.

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

1984-10-09T23:59:59.000Z

293

Lithium aluminate/zirconium material useful in the production of tritium  

DOE Patents [OSTI]

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.

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

1984-10-09T23:59:59.000Z

294

Brookhaven National Laboratory - HFBR Tritium | 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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of Energy Power Systems EngineeringDepartment of4 Federal6Clean Energy | DepartmentBrandonYoungHFBR Tritium

295

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

E-Print Network [OSTI]

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

Williams, Roy

296

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

E-Print Network [OSTI]

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

Aeschbach-Hertig, Werner

297

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

E-Print Network [OSTI]

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

Basse, Nils Plesner

298

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 for tritium-suppressed D-D fusion and the understanding of the turbulent pinch in magnetically confined plasma pathway. Tritium- suppressed D-D fusion eliminates the need to breed fuel from lithium, reduces the damage

299

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

E-Print Network [OSTI]

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

Harilal, S. S.

300

Mixed Waste Management Facility groundwater monitoring report. Second quarter 1994  

SciTech Connect (OSTI)

Currently, 125 wells monitor groundwater quality in the uppermost aquifer beneath the Mixed Waste Management Facility (MWMF) at the Savannah River Site. Samples from the wells are analyzed for selected heavy metals, indicator parameters, radionuclides, volatile organic compounds, and other constituents. During second quarter 1994, chloroethene (vinyl chloride), 1,1-dichloroethylene, gross alpha, lead, tetrachloroethylene, trichloroethylene, or tritium exceeded final Primary Drinking Water Standards (PDWS) in approximately half of the downgradient wells at the MWMF. Consistent with historical trends, elevated constituent levels were found primarily in Aquifer Zone. As in previous quarters, tritium and trichloroethylene were the most widespread elevated constituents during second quarter 1994. Sixty-two of the 125 monitoring wells contained elevated tritium activities. Trichloroethylene concentrations exceeded the final PDWS in 23 wells. Chloroethene, 1,1-dichloroethylene, lead, and tetrachloroethylene, elevated in one or more wells during second quarter 1994, also occurred in elevated levels during first quarter 1994. These constituents generally were elevated in the same wells during both quarters. Gross alpha, which was not elevated in any well during first quarter 1994, was elevated in one well during second quarter. Copper, mercury, and nonvolatile beta were elevated during first quarter 1994 but not during second quarter.

Chase, J.A.

1994-09-01T23:59:59.000Z

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


301

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

SciTech Connect (OSTI)

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.

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

1995-10-01T23:59:59.000Z

302

PUREX facility hazards assessment  

SciTech Connect (OSTI)

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.

Sutton, L.N.

1994-09-23T23:59:59.000Z

303

Accelerator Production of Tritium project process waste assessment  

SciTech Connect (OSTI)

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).

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

1995-09-01T23:59:59.000Z

304

Radiological performance assessment for the E-Area Vaults Disposal Facility  

SciTech Connect (OSTI)

The E-Area Vaults (EAVs) located on a 200 acre site immediately north of the current LLW burial site at Savannah River Site will provide a new disposal and storage site for solid, low-level, non-hazardous radioactive waste. The EAV Disposal Facility will contain several large concrete vaults divided into cells. Three types of structures will house four designated waste types. The Intermediate Level Non-Tritium Vaults will receive waste radiating greater than 200 mR/h at 5 cm from the outer disposal container. The Intermediate Level Tritium Vaults will receive waste with at least 10 Ci of tritium per package. These two vaults share a similar design, are adjacent, share waste handling equipment, and will be closed as one facility. The second type of structure is the Low Activity Waste Vaults which will receive waste radiating less than 200 mR/h at 5 cm from the outer disposal container and containing less than 10 Ci of tritium per package. The third facility, the Long Lived Waste Storage Building, provides covered, long term storage for waste containing long lived isotopes. Two additional types of disposal are proposed: (1) trench disposal of suspect soil, (2) naval reactor component disposal. To evaluate the long-term performance of the EAVs, site-specific conceptual models were developed to consider: (1) exposure pathways and scenarios of potential importance; (2) potential releases from the facility to the environment; (3) effects of degradation of engineered features; (4) transport in the environment; (5) potential doses received from radionuclides of interest in each vault type.

Cook, J.R.; Hunt, P.D. [Westinghouse Savannah River Co., Aiken, SC (United States)

1994-04-15T23:59:59.000Z

305

Coal extraction  

SciTech Connect (OSTI)

Coal is extracted using a mixed solvent which includes a substantially aromatic component and a substantially naphthenic component, at a temperature of 400/sup 0/ to 500/sup 0/C. Although neither component is an especially good solvent for coal by itself, the use of mixed solvent gives greater flexibility to the process and offers efficiency gains.

Clarke, J.W.; Kimber, G.M.; Rantell, T.D.; Snape, C.E.

1985-06-04T23:59:59.000Z

306

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

SciTech Connect (OSTI)

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

Chang H. Oh; Eung S. Kim

2009-09-01T23:59:59.000Z

307

Facilities Services Overview & Discussion  

E-Print Network [OSTI]

& Finance Facilities Services Director: Jeff Butler Human Resources Administrative Services Engineering) Environmental Services Morrison (3) Admin Services Evans (1) Human Resources Engineering (4) ·EngineeringFacilities Services Overview & Discussion Jeff Butler Director ­ Facilities Services November 2011

Maxwell, Bruce D.

308

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

E-Print Network [OSTI]

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.

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

2013-06-03T23:59:59.000Z

309

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

E-Print Network [OSTI]

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.

Barmina, E V; Timashev, S F; Shafeev, G A

2013-01-01T23:59:59.000Z

310

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

SciTech Connect (OSTI)

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.

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

1988-01-01T23:59:59.000Z

311

Hydrodynamic instability growth and mix experiments at the National Ignition Facility  

SciTech Connect (OSTI)

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.

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

312

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...

313

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...

314

Guide to research facilities  

SciTech Connect (OSTI)

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.

Not Available

1993-06-01T23:59:59.000Z

315

Assessment of radiation exposure for materials in the LANSCE Spallation Irradiation Facility  

SciTech Connect (OSTI)

Materials samples were irradiated in the Los Alamos Radiation Effects Facility (LASREF) at the Los Alamos Neutron Science Center (LANSCE) to provide data for the Accelerator Production of Tritium (APT) project on the changes in mechanical and physical properties of materials in a spallation target environment. The targets were configured to expose samples to a variety of radiation environments including high-energy protons, mixed protons and neutrons, and predominantly neutrons. The irradiation was driven by an 800 MeV 1 mA proton beam with a circular Gaussian shape of approximately 2{sigma} = 3.5 cm. Two irradiation campaigns were conducted in which samples were exposed for approximately six months and two months, respectively. At the end of this period, the samples were extracted and tested. Activation foils that had been placed in proximity to the materials samples were used to quantify the fluences in various locations. The STAYSL2 code was used to estimate the fluences by combining the activation foil data with calculated data from the LAHET Code System (LCS) and MCNPX. The exposure for each sample was determined from the estimated fluences using interpolation based on a mathematical fitting to the fluence results. The final results included displacement damage (dpa) and gas (H, He) production for each sample from the irradiation. Based on the activation foil analysis, samples from several locations in both irradiation campaigns were characterized. The radiation damage to each sample was highly dependent upon location and varied from 0.023 to 13 dpa and was accompanied by high levels of H and He production.

James, M. R. (Michael R.); Maloy, S. A. (Stuart A.); Sommer, W. F. (Walter F.), Jr.; Fowler, Malcolm M.; Dry, D. E. (Donald E.); Ferguson, P. D. (Phillip D.); Corzine, R. K. (R. Karen); Mueller, G. E. (Gary E.)

2001-01-01T23:59:59.000Z

316

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

SciTech Connect (OSTI)

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.

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

1995-03-01T23:59:59.000Z

317

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

SciTech Connect (OSTI)

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

Warren, R.

2012-10-08T23:59:59.000Z

318

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

319

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

320

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

E-Print Network [OSTI]

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 ...

Rosenberg, Michael Jonathan

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


321

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

SciTech Connect (OSTI)

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.

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

322

A Study on the Tritium Behavior in the Rice Plant after a Short-Term Exposure of HTO  

SciTech Connect (OSTI)

In many Asian countries including Korea, rice is a very important food crop. Its grain is consumed by humans and its straw is used to feed animals. In Korea, there are four CANDU type reactors that release relatively large amounts of tritium into the environment. Since 1997, KAERI (Korea Atomic Energy Research Institute) has carried out the experimental studies to obtain domestic data on various parameters concerning the direct contamination of plant. In this study, the behavior of tritium in the rice plant is predicted and compared with the measurement performed at KAERI. Using the conceptual model of the soil-plant-atmosphere tritiated water transport system which was suggested by Charles E. Murphy, tritium concentrations in the soil and in leaves to time were derived. If the effect of tritium concentration in the soil is considered, the tritium concentration in leaves is described as a double exponential model. On the other hand if the tritium concentration in the soil is disregarded, the tritium concentration in leaves is described by a single exponential term as other models (e.g. Belot's or STAR-H3 model). Also concentration of organically bound tritium in the seed is predicted and compared with measurements. The results can be used to predict the tritium concentration in the rice plant at a field around the site and the ingestion dose following the release of tritium to the environment.

Yook, D-S.; Lee, K. J.; Choi, Y-H.

2002-02-26T23:59:59.000Z

323

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

SciTech Connect (OSTI)

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.

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

2011-05-01T23:59:59.000Z

324

Future Fixed Target Facilities  

SciTech Connect (OSTI)

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.

Melnitchouk, Wolodymyr

2009-01-01T23:59:59.000Z

325

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

SciTech Connect (OSTI)

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.

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

326

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

SciTech Connect (OSTI)

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.

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

327

MANAGING BERYLLIUM IN NUCLEAR FACILITY APPLICATIONS  

SciTech Connect (OSTI)

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.

R. Rohe; T. N. Tranter

2011-12-01T23:59:59.000Z

328

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

SciTech Connect (OSTI)

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)

Not Available

1988-12-01T23:59:59.000Z

329

Tritium distribution in the environment in the vicinity of a chronic atmospheric source-assessment of the steady state hypothesis  

SciTech Connect (OSTI)

The Savannah River Site (SRS) is a major radionuclide production center. Tritium has been released to the atmosphere over the 36 year period of operation. The tritiated water concentration of the atmosphere, rain, vegetation and food have been routinely monitored during this period. Special studies have been made of tritium in soils and in the organic fractions of these same materials. The available data suggest that the average tritium concentration in the components of the terrestrial environment have approached a steady state with the two main sources of tritium, rainfall and atmospheric water vapor.

Murphy, C.E. Jr.; Bauer, L.R.; Zeigler, C.C.

1990-01-01T23:59:59.000Z

330

Tritium distribution in the environment in the vicinity of a chronic atmospheric source-assessment of the steady state hypothesis  

SciTech Connect (OSTI)

The Savannah River Site (SRS) is a major radionuclide production center. Tritium has been released to the atmosphere over the 36 year period of operation. The tritiated water concentration of the atmosphere, rain, vegetation and food have been routinely monitored during this period. Special studies have been made of tritium in soils and in the organic fractions of these same materials. The available data suggest that the average tritium concentration in the components of the terrestrial environment have approached a steady state with the two main sources of tritium, rainfall and atmospheric water vapor.

Murphy, C.E. Jr.; Bauer, L.R.; Zeigler, C.C.

1990-12-31T23:59:59.000Z

331

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

SciTech Connect (OSTI)

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.

Shaver, Mark W.; Lanning, Donald D.

2010-02-01T23:59:59.000Z

332

Non-conventional passive sensors for monitoring tritium on surfaces  

SciTech Connect (OSTI)

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.

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

333

Tritium control and activation in the Pulse*Star reactor  

SciTech Connect (OSTI)

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.

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

1983-12-01T23:59:59.000Z

334

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.

335

Phase 1 Final Report for In-Situ Tritium Beta Detector  

SciTech Connect (OSTI)

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.

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

1998-04-15T23:59:59.000Z

336

FACILITY SAFETY (FS)  

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

FACILITY SAFETY (FS) OBJECTIVE FS.1 - (Core Requirement 7) Facility safety documentation in support of SN process operations,is in place and has been implemented that describes the...

337

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...

338

Apparatus for hydrocarbon extraction  

DOE Patents [OSTI]

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.

Bohnert, George W.; Verhulst, Galen G.

2013-03-19T23:59:59.000Z

339

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

SciTech Connect (OSTI)

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.

Hysong, R. J.

1998-12-21T23:59:59.000Z

340

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

SciTech Connect (OSTI)

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.

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

1995-09-01T23:59:59.000Z

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


341

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

SciTech Connect (OSTI)

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)

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

342

Extraction efficiency and quantification of mutagenic chemicals in soils  

E-Print Network [OSTI]

: 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...

Maggard, Lea Ann

1986-01-01T23:59:59.000Z

343

Better building: LEEDing new facilities  

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

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

344

Development of Tritium AMS for Biomedical Sciences Research  

SciTech Connect (OSTI)

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.

Dingley, K H; Chiarappa-Zucca, M L

2002-01-01T23:59:59.000Z

345

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

346

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

347

Migration in alluvium of chlorine-36 and tritium from an underground nuclear test  

SciTech Connect (OSTI)

This article describes a field experiment studying the migration in alluvium of radioactive elements away from an underground nuclear explosion at the Nevada Test Site in the United States. Nuclides detected in the pumped water are tritium, chlorine-36, iodine-129, and krypton-85 - all at levels below the maximum permissible concentration for drinking water in controlled areas. The chlorine-36 elution curve precedes that of tritium, and is due to an anion exclusion process. A conventional two-dimensional convection-diffusion equation does not fully describe the elution curves for tritium and chlorine-36; the tailing of the curves is longer than predicted. Successful modeling of this experiment will be important for validating codes and models to be used in the high-level nuclear waste program.

Ogard, A.E.; Thompson, J.L.; Rundberg, R.S.; Wolfsberg, K.; Kubik, P.W.; Elmore, D.; Bentley, H.W.

1987-01-01T23:59:59.000Z

348

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

E-Print Network [OSTI]

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.

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

2012-09-24T23:59:59.000Z

349

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

E-Print Network [OSTI]

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.

,

2013-01-01T23:59:59.000Z

350

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

SciTech Connect (OSTI)

'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.'

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

1997-09-01T23:59:59.000Z

351

Radioactive Liquid Waste Treatment Facility Discharges in 2011  

SciTech Connect (OSTI)

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

Del Signore, John C. [Los Alamos National Laboratory

2012-05-16T23:59:59.000Z

352

Mixed Waste Management Facility groundwater monitoring report: Third quarter 1994  

SciTech Connect (OSTI)

Currently, 125 wells monitor groundwater quality in the uppermost aquifer beneath the Mixed Waste Management Facility (MWMF) at the Savannah River Site. Samples from the wells are analyzed for selected heavy metals, herbicides/pesticides, indicator parameters, radionuclides, volatile organic compounds, and other constituents. As in previous quarters, tritium and trichloroethylene were the most widespread elevated constituents during third quarter 1994. Sixty-four (51%) of the 125 monitoring wells contained elevated tritium activities. Trichloroethylene concentrations exceeded the final PDWS in 22 (18%) wells. Chloroethene, 1,1-dichloroethylene, and tetrachloroethylene, elevated in one or more wells during third quarter 1994, also occurred in elevated levels during second quarter 1994. These constituents generally were elevated in the same wells during both quarters. Gross alpha, which was elevated in only one well during second quarter 1994, was elevated again during third quarter. Mercury, which was elevated during first quarter 1994, was elevated again in one well. Dichloromethane was elevated in two wells for the first time in several quarters.

Not Available

1994-12-01T23:59:59.000Z

353

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

E-Print Network [OSTI]

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

Aeschbach-Hertig, Werner

354

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

355

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

SciTech Connect (OSTI)

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.

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

356

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

SciTech Connect (OSTI)

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

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

2002-05-03T23:59:59.000Z

357

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

SciTech Connect (OSTI)

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.

R. A. Causey

1999-02-01T23:59:59.000Z

358

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:...

359

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

SciTech Connect (OSTI)

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.

Nichols, R.L.

1993-03-03T23:59:59.000Z

360

Stockpile Stewardship and the National Ignition Facility  

SciTech Connect (OSTI)

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.

Moses, E

2012-01-04T23:59:59.000Z

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


361

Mixed Waste Management Facility groundwater monitoring report, First quarter 1994  

SciTech Connect (OSTI)

During first quarter 1994, nine constituents exceeded final Primary Drinking Water Standards in groundwater samples from downgradient monitoring wells at the Mixed Waste Management Facility, the Old Burial Ground, the E-Area Vaults, the proposed Hazardous Waste/Mixed Waste Disposal Vaults, and the F-Area Sewage Sludge Application Site. As in previous quarters, tritium and trichloroethylene were the most widespread elevated constituents. Chloroethene (vinyl chloride), copper, 1,1-dichloroethylene, lead, mercury, nonvolatile beta, or tetrachloroethylene also exceeded standards in one or more wells. Elevated constituents were found in numerous Aquifer Zone IIB{sub 2} (Water Table) and Aquifer Zone IIB{sub 1}, (Barnwell/McBean) wells and in one Aquifer Unit IIA (Congaree) well. The groundwater flow directions and rates in the three hydrostratigraphic units were similar to those of previous quarters.

Not Available

1994-06-01T23:59:59.000Z

362

Inertial Confinement Fusion and the National Ignition Facility (NIF)  

SciTech Connect (OSTI)

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.

Ross, P.

2012-08-29T23:59:59.000Z

363

Waste Management facilities fault tree databank 1995 status report  

SciTech Connect (OSTI)

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.

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

1995-08-16T23:59:59.000Z

364

Science and Technology Facility  

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

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

365

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...

366

FACILITY SAFETY (FS)  

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

- (Core Requirements 4 and 6) Sufficient numbers of qualified personnel are available to conduct and support operations. Adequate facilities and equipment are available to ensure...

367

ARM Mobile Facilities  

ScienceCinema (OSTI)

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.

Orr, Brad; Coulter, Rich

2014-09-15T23:59:59.000Z

368

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...

369

Idaho National Laboratory Facilities  

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

National Scientific User Facility Center for Advanced Energy Studies Light Water Reactor Sustainability Idaho Regional Optical Network LDRD Next Generation Nuclear Plant Docs...

370

Supercomputing | Facilities | ORNL  

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

facilities, and authorization checks for physical access. An integrated cyber security plan encompasses all aspects of computing. Cyber security plans are risk-based....

371

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.

372

Hot Fuel Examination Facility  

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

Working with INL Community Outreach Visitor Information Calendar of Events ATR National Scientific User Facility Center for Advanced Energy Studies Light Water Reactor...

373

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)...

374

Mixed Waste Management Facility (MWMF) groundwater monitoring report. Second quarter 1993  

SciTech Connect (OSTI)

Groundwater monitoring continued at the Savannah River Plant. During second quarter 1993, nine constituents exceeded final Primary Drinking Water Standards in groundwater samples from downgradient monitoring wells at the Mixed Waste Management Facility, the Old Burial Ground, the E-Area Vaults, and the proposed Hazardous Waste/Mixed Waste Disposal Vaults. As in previous quarters, tritium and trichloroethylene were the most widespread constituents. Chloroethene (vinyl chloride), dichloromethane (methylene chloride), 1,1-dichloroethylene, gross alpha, lead, nonvolatile beta, or tetrachloroethylene also exceeded standards in one or more wells. The groundwater flow directions and rates in the three hydrostratigraphic units were similar to those of previous quarters.

Not Available

1993-09-01T23:59:59.000Z

375

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

SciTech Connect (OSTI)

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.

Not Available

1993-09-30T23:59:59.000Z

376

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

SciTech Connect (OSTI)

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.

Atkinson, R.

2012-07-31T23:59:59.000Z

377

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

SciTech Connect (OSTI)

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.

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

378

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

SciTech Connect (OSTI)

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.

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

379

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

SciTech Connect (OSTI)

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.

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

2012-07-01T23:59:59.000Z

380

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

E-Print Network [OSTI]

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_\

Andrew J. Long; Cecilia Lunardini; Eray Sabancilar

2014-11-12T23:59:59.000Z

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


381

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

E-Print Network [OSTI]

a b s t r a c t In order to measure groundwater age and design nuclear waste disposal sites, both of which could be considered candidate sites for Very Low Level Waste disposal; silty sand from to be considered in fate and transport studies of tritium in the environment. Ă? 2010 Elsevier Ltd. All rights

Hu, Qinhong "Max"

382

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

E-Print Network [OSTI]

, 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

Abdou, Mohamed

383

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

384

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

385

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

SciTech Connect (OSTI)

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.

Chang H. Oh; Eung S. Kim

2008-09-01T23:59:59.000Z

386

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

E-Print Network [OSTI]

the best itvailable copy. Available lo I)()E and I)()E Ct)lllritclc)_"st"1"+,)111the ()rf ice of Scientific into a lithium target to produce

Harilal, S. S.

387

Fusion Nuclear Schience Facility-AT: A Material And Component Testing Device  

SciTech Connect (OSTI)

A Fusion Nuclear Science Facility (FNSF) is a necessary complement to ITER, especially in the area of materials and components testing, needed for DEMO design development. FNSF-AT, which takes advantage of advanced tokamak (AT) physics should have neutron wall loading of 1-2 MW/m2, continuous operation for periods of up to two weeks, a duty factor goal of 0.3 per year and an accumulated fluence of 3-6 MW-yr/m2 (~30-60 dpa) in ten years to enable the qualification of structural, blanket and functional materials, components and corresponding ancillary equipment necessary for the design and licensing of a DEMO. Base blankets with a ferritic steel structure and selected tritium blanket materials will be tested and used for the demonstration of tritium sufficiency. Additional test ports at the outboard mid-plane will be reserved for test blankets with advanced designs or exotic materials, and electricity production for integrated high fluence testing in a DT fusion spectrum. FNSF-AT will be designed using conservative implementations of all elements of AT physics to produce 150-300 MW fusion power with modest energy gain (Q<7) in a modest sized normal conducting coil device. It will demonstrate and help to select the DEMO plasma facing, structural, tritium breeding, functional materials and ancillary equipment including diagnostics. It will also demonstrate the necessary tritium fuel cycle, design and cooling of the first wall chamber and divertor components. It will contribute to the knowledge on material qualification, licensing, operational safety and remote maintenance necessary for DEMO design

Wong, C. P.; Chan, V. S.; Garofalo, A. M.; Stambaugh, Ron; Sawan, M.; Kurtz, Richard J.; Merrill, Brad

2012-07-01T23:59:59.000Z

388

Privacy Impact Assessment OFEO Facilities Management System Facilities Center  

E-Print Network [OSTI]

Privacy Impact Assessment OFEO Facilities Management System ­ Facilities Center I. System Identification 1. IT System Name: Facilities Management System - FacilityCenter 2. IT System Sponsor: Office. IT System Manager: Michelle T. Gooch, Facilities Management Systems Manager 5. PIA Author: Michelle T. Gooch

Mathis, Wayne N.

389

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

SciTech Connect (OSTI)

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.

C. Cooper; M. Ye; J. Chapman

2008-04-01T23:59:59.000Z

390

Facilities Management CAD Standards  

E-Print Network [OSTI]

Facilities Management CAD Standards 2011 #12;Facilities Management CAD Standards Providing: Layering Standards 2.1 Layer Name Format 2.2 Layer Name Modifiers 2.3 Layer Attributes 2.4 Special Layer of PDF and DWG Files APPENDIX A: DAL FM CAD Standard Layers APPENDIX B: DAL FM CAD Special Layers

Brownstone, Rob

391

Cornell University Facilities Services  

E-Print Network [OSTI]

requirements, building code, and sustainability objectives. This plan takes a long- term view, projecting workCornell University Facilities Services Contract Colleges Facilities Fernow and Rice Hall in Fernow, Rice, Bruckner, Bradfield and Plant Science buildings. It includes a surging and phasing plan

Manning, Sturt

392

Argonne Leadership Computing Facility  

E-Print Network [OSTI]

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

Kemner, Ken

393

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

394

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

395

Science &Technology Facilities Council  

E-Print Network [OSTI]

and Science & Technology Facilities Council invite you to The ESA Technology Transfer Network SpaceTech2012Science &Technology Facilities Council Innovations Issue 31 October 2012 This issue: 1 STFC International prize for `no needles' breast cancer diagnosis technique 6 CEOI Challenge Workshop ­ Current

396

Emergency Facilities and Equipment  

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

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.

1997-08-21T23:59:59.000Z

397

Catalytic extraction processing of contaminated scrap metal  

SciTech Connect (OSTI)

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.

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

1995-12-01T23:59:59.000Z

398

Thermal-hydraulic design of the target/blanket for the accelerator production of tritium conceptual design  

SciTech Connect (OSTI)

A conceptual design was developed for the target/blanket system of an accelerator-based system to produce tritium. The target/blanket system uses clad tungsten rods for a spallation target and clad lead rods as a neutron multiplier in a blanket surrounding the target. The neutrons produce tritium in {sup 3}He, which is contained in aluminum tubes located in the decoupler and blanket regions. This paper presents the thermal-hydraulic design of the target, decoupler, and blanket developed for the conceptual design of the Accelerator Production of Tritium Project, and demonstrates there is adequate margin in the design at full power operation.

Willcutt, G.J.E. Jr.; Kapernick, R.J.

1997-11-01T23:59:59.000Z

399

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

SciTech Connect (OSTI)

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)

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

400

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

SciTech Connect (OSTI)

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).

Sullivan, N.

1995-05-02T23:59:59.000Z

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


401

Department of Residential Facilities Facilities Student Employment Office  

E-Print Network [OSTI]

Department of Residential Facilities Facilities Student Employment Office 1205E Leonardtown Service Updated 3/09 #12;EMPLOYMENT HISTORY Have you worked for Residential Facilities before? Yes No If so list

Hill, Wendell T.

402

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...

403

Information extraction system  

DOE Patents [OSTI]

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.

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

404

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

E-Print Network [OSTI]

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 ...

Goncharov, V. N.

405

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

E-Print Network [OSTI]

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 ...

Rodriguez, Judy N

2013-01-01T23:59:59.000Z

406

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

SciTech Connect (OSTI)

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.

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

2012-02-19T23:59:59.000Z

407

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,...

408

NETL - Fuel Reforming Facilities  

ScienceCinema (OSTI)

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.

None

2014-06-27T23:59:59.000Z

409

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

410

NEW RENEWABLE FACILITIES PROGRAM  

E-Print Network [OSTI]

CALIFORNIA ENERGY COMMISSION NEW RENEWABLE FACILITIES PROGRAM GUIDEBOOK APRIL 2006 CEC-300 Director Heather Raitt Technical Director Renewable Energy Program Drake Johnson Office Manager Renewable Energy Office Valerie Hall Deputy Director Efficiency, Renewables, and Demand Analysis Division #12;These

411

Liquidity facilities and signaling  

E-Print Network [OSTI]

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 ...

Arregui, Nicolás

2010-01-01T23:59:59.000Z

412

NETL - Fuel Reforming Facilities  

SciTech Connect (OSTI)

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.

None

2013-06-12T23:59:59.000Z

413

Cornell University Facilities Services  

E-Print Network [OSTI]

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

Manning, Sturt

414

B Plant facility description  

SciTech Connect (OSTI)

Buildings 225B, 272B, 282B, 282BA, and 294B were removed from the B Plant facility description. Minor corrections were made for tank sizes and hazardous and toxic inventories.

Chalk, S.E.

1996-10-04T23:59:59.000Z

415

Facilities Management Department Restructuring  

E-Print Network [OSTI]

­ 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

Mullins, Dyche

416

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.

417

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...

418

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...

419

Facility Operations 1993 fiscal year work plan: WBS 1.3.1  

SciTech Connect (OSTI)

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).

Not Available

1992-11-01T23:59:59.000Z

420

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

E-Print Network [OSTI]

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 ...

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

2012-01-01T23:59:59.000Z

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


421

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

SciTech Connect (OSTI)

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.

Clark, E; Marie Kane, M

2008-12-12T23:59:59.000Z

422

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

SciTech Connect (OSTI)

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.

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

2012-01-01T23:59:59.000Z

423

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

SciTech Connect (OSTI)

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.

Mills, Bernice E.

2007-11-01T23:59:59.000Z

424

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

E-Print Network [OSTI]

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.

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

425

Hanford Site Near-Facility Environmental Monitoring Data Report for Calendar Year 2008  

SciTech Connect (OSTI)

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.

Perkins, Craig J.; Dorsey, Michael C.; Mckinney, Stephen M.; Wilde, Justin W.; Poston, Ted M.

2009-09-15T23:59:59.000Z

426

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

SciTech Connect (OSTI)

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.

Carsten, A.L.

1984-01-01T23:59:59.000Z

427

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

E-Print Network [OSTI]

-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...

Fryar, Alan Ernest

1986-01-01T23:59:59.000Z

428

Tritium and helium analyses in thin films by enhanced proton backscattering  

E-Print Network [OSTI]

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.

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

2013-10-14T23:59:59.000Z

429

Fuel assembly for the production of tritium in light water reactors  

DOE Patents [OSTI]

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.

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

1985-01-01T23:59:59.000Z

430

Fuel assembly for the production of tritium in light water reactors  

DOE Patents [OSTI]

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.

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

1983-06-10T23:59:59.000Z

431

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

SciTech Connect (OSTI)

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.

D. SIEBE; K. PASAMEHMETOGLU

2000-11-01T23:59:59.000Z

432

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

SciTech Connect (OSTI)

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.

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

433

Working with SRNL - Our Facilities - Glovebox Facilities  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear SecurityTensile Strain Switched FerromagnetismWaste and MaterialsWenjun1 Table 1.14Working WithGlovebox Facilities

434

Brookhaven Facility Biomass Facility | Open Energy Information  

Open Energy Info (EERE)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home5b9fcbce19 NoPublic Utilities Address: 160Benin: EnergyBoston Areais a village in Cook County, Illinois. ItBrookhaven Facility

435

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

SciTech Connect (OSTI)

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.

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

436

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

E-Print Network [OSTI]

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.

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

2014-05-15T23:59:59.000Z

437

DESIGN OF A CONTAINMENT VESSEL CLOSURE FOR SHIPMENT OF TRITIUM GAS  

SciTech Connect (OSTI)

This paper presents a design summary of the containment vessel closure for the Bulk Tritium Shipping Package (BTSP). This new package is 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 new design is based on changes in the regulatory requirements. 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 that enhance the package's ability to withstand dynamic loading and thermal effects. This paper will specifically summarize the design philosophy and engineered features of the BTSP containment vessel closure. The closure design incorporates a concave closure lid, metallic C-Ring seals for containing tritium gas, a metal bellows valve and an elastomer O-Ring for leak testing. The efficient design minimizes the overall vessel height and protects the valve housing from damage during postulated drop and crush scenarios. Design features will be discussed.

Eberl, K; Paul Blanton, P

2007-07-03T23:59:59.000Z

438

UNIVERSITY BOULEVARD FAU Research Facility  

E-Print Network [OSTI]

Harriet L.Wilkes Honors College FAU Research Facility Expansion Satellite Utility Plant Chiller Lift

Fernandez, Eduardo

439

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

E-Print Network [OSTI]

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

Paris-Sud XI, Université de

440

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]

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

Paris-Sud XI, Université de

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


441

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

E-Print Network [OSTI]

959 PRODUCTION DE TRITIUM DANS LE THORIUM PAR DES PROTONS DE 135 MeV Par M. LEFORT, G. SIMONOFF et Saclay. Résumé. 2014 On a mesuré la section efficace de production de tritium après bombardment de thorium par des protons de 135 MeV accélérés au synchro-cyclotron d'Orsay. Le tritium était extrait des

Paris-Sud XI, Université de

442

Hanford facility contingency plan  

SciTech Connect (OSTI)

The Hanford Facility Contingency Plan, together with each TSD unit- specific contingency plan, meets the WAC 173-303 requirements for a contingency plan. Applicability of this plan to Hanford Facility activities is described in the Hanford Facility RCRA Permit, Dangerous Waste Portion, General Condition II.A. General Condition II.A applies to Part III TSD units, Part V TSD units, and to releases of hazardous substances which threaten human health or the environment. Additional information about the applicability of this document may also be found in the Hanford Facility RCRA Permit Handbook (DOE/RL-96-10). This plan includes descriptions of responses to a nonradiological hazardous substance spill or release at Hanford Facility locations not covered by TSD unit-specific contingency plans or building emergency plans. The term hazardous substances is defined in WAC 173-303-040 as: ``any liquid, solid, gas, or sludge, including any material, substance, product, commodity, or waste, regardless of quantity, that exhibits any of the physical, chemical or biological properties described in WAC 173-303-090 or 173-303-100.`` Whenever the term hazardous substances is used in this document, it will be used in the context of this definition. This plan includes descriptions of responses for spills or releases of hazardous substances occurring at areas between TSD units that may, or may not, threaten human health or the environment.

Sutton, L.N.

1996-07-01T23:59:59.000Z

443

National Ignition Facility Project Completion and Control System Status  

SciTech Connect (OSTI)

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.

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

444

Regional groundwater flow and tritium transport modeling and risk assessment of the underground test area, Nevada Test Site, Nevada  

SciTech Connect (OSTI)

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

None

1997-10-01T23:59:59.000Z

445

Fission Product Extraction Process  

SciTech Connect (OSTI)

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

None

2011-01-01T23:59:59.000Z

446

Coal extraction process  

SciTech Connect (OSTI)

Sub-divided coal is extracted under non-thermally destructive conditions with a solvent liquid containing a compound having the general formula:

Hammack, R. W.; Sears, J. T.; Stiller, A. H.

1981-06-09T23:59:59.000Z

447

Fission Product Extraction Process  

ScienceCinema (OSTI)

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

None

2013-05-28T23:59:59.000Z

448

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

449

Fitness facilities, facilities for extracurricular activities and other purposes Facility Location Department in charge  

E-Print Network [OSTI]

Facility Location Department in charge Student Hall (1) Common Facility 1 for Extracurricular Activities (2 tennis courts, Swimming pool (25 m, not officially approved) Rokkodai Area (Tsurukabuto 2 Campus) Martial art training facility, Japanese archery training facility, Playground, 4 tennis courts, Swimming pool

Banbara, Mutsunori

450

RCRA facility stabilization initiative  

SciTech Connect (OSTI)

The RCRA Facility Stabilization Initiative was developed as a means of implementing the Corrective Action Program`s management goals recommended by the RIS for stabilizing actual or imminent releases from solid waste management units that threaten human health and the environment. The overall goal of stabilization is to, as situations warrant, control or abate threats to human health and/or the environment from releases at RCRA facilities, and/or to prevent or minimize the further spread of contamination while long-term remedies are pursued. The Stabilization initiative is a management philosophy and should not be confused with stabilization technologies.

Not Available

1995-02-01T23:59:59.000Z

451

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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsrucLasDelivered‰PNGExperience4AJ01)3, 2010September 30, 2004 [FacilityMayMarch 1, 2012 [Facility

452

ARM - Facility News Article  

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

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453

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:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series toESnet4:Epitaxial ThinFOR IMMEDIATE5Facilities SomeFacilities Glove

454

Service & Reliability Equipment & Facilities  

E-Print Network [OSTI]

termites E5 Marine applications, panel & block E7 Field Stake tests (FST colonies) E9 Above ground L-joint stake test (Formosan termites & decay), E9 L- joint, E16 (horizontal lap-joint), E18 ground proximity facilities for AWPA test: A9 X-ray, E1 (termites), E10 (soil block), E11 (leaching), E12 metal corrosion

455

Graph algorithms experimentation facility  

E-Print Network [OSTI]

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...

Sonom, Donald George

1994-01-01T23:59:59.000Z

456

Strategies for Facilities Renewal  

E-Print Network [OSTI]

of steam production is from exothermic chem ical processes. A large gas fired cogeneration unit was completed in 1987 and supplies 90% of the facil ities' electrical needs and 25% of total steam demand (the remaining steam is supplied by process heat...

Good, R. L.

457

FACILITIES INSTRUCTIONS, STANDARDS, & TECHNIQUES  

E-Print Network [OSTI]

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

Laughlin, Robert B.

458

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.

459

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

460

The Caterpillar Coal Gasification Facility  

E-Print Network [OSTI]

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...

Welsh, J.; Coffeen, W. G., III

1983-01-01T23:59:59.000Z

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


461

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

462

Sensitivity of Next-Generation Tritium Beta-Decay Experiments for keV-Scale Sterile Neutrinos  

E-Print Network [OSTI]

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.

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

463

Exact relativistic tritium beta-decay endpoint spectrum in a hadron model  

E-Print Network [OSTI]

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.

Fedor Simkovic; Rastislav Dvornicky; Amand Faessler

2008-05-05T23:59:59.000Z

464

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

SciTech Connect (OSTI)

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.

None

1996-12-01T23:59:59.000Z

465

Recoilless Resonance Absorption of Tritium Antineutrinos and Time-Energy Uncertainty Relation  

E-Print Network [OSTI]

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.

S. M. Bilenky

2007-08-02T23:59:59.000Z

466

Facilities evaluation report  

SciTech Connect (OSTI)

The Buried Waste Integrated Demonstration (BWID) is a program of the Department of Energy (DOE) Office of Technology Development whose mission is to evaluate different new and existing technologies and determine how well they address DOE community waste remediation problems. Twenty-three Technical Task Plans (TTPs) have been identified to support this mission during FY-92; 10 of these have identified some support requirements when demonstrations take place. Section 1 of this report describes the tasks supported by BWID, determines if a technical demonstration is proposed, and if so, identifies the support requirements requested by the TTP Principal Investigators. Section 2 of this report is an evaluation identifying facility characteristics of existing Idaho National Engineering Laboratory (INEL) facilities that may be considered for use in BWID technology demonstration activities.

Sloan, P.A.; Edinborough, C.R.

1992-04-01T23:59:59.000Z

467

PUREX facility preclosure work plan  

SciTech Connect (OSTI)

This preclosure work plan presents a description of the PUREX Facility, the history of the waste managed, and addresses transition phase activities that position the PUREX Facility into a safe and environmentally secure configuration. For purposes of this documentation, the PUREX Facility does not include the PUREX Storage Tunnels (DOE/RL-90/24). Information concerning solid waste management units is discussed in the Hanford Facility Dangerous Waste Permit Application, General Information Portion (DOE/RL-91-28, Appendix 2D).

Engelmann, R.H.

1997-04-24T23:59:59.000Z

468

Reed Reactor Facility Annual Report  

SciTech Connect (OSTI)

This is the report of the operations, experiments, modifications, and other aspects of the Reed Reactor Facility for the year.

Frantz, Stephen G.

2000-09-01T23:59:59.000Z

469

Lunch & Learn Facilities &  

E-Print Network [OSTI]

" 3 #12;What are F&A costs? OMB Circular A-21 provides guidance on F&A costs F&A a.k.a. Overhead a #12;F&A Rate Development Process FSU's process must be designed to ensure that Federal sponsors do usage ­ Allocate facilities costs ­ Provide productivity analysis Space survey tool WebSpace ­ On-line

McQuade, D. Tyler

470

ARM - SGP Intermediate Facility  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsruc Documentation RUC :ProductsSCM Forcing Data DerivedInstrumentsPolarExtended Facility

471

Facilities | Department of Energy  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE: Alternative Fuels DataCombined Heat & PowerEnergy BlogExchangeSummary TableFacilities

472

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...

473

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

474

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...

475

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...

476

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)

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.

Dixon, K.L.; Rogers, V.A.; Looney, B.B.

1994-06-01T23:59:59.000Z

477

Liquid chromatographic extraction medium  

DOE Patents [OSTI]

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.

Horwitz, E. Philip (Naperville, IL); Dietz, Mark L. (Evanston, IL)

1994-01-01T23:59:59.000Z

478

Liquid chromatographic extraction medium  

DOE Patents [OSTI]

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.

Horwitz, E.P.; Dietz, M.L.

1994-09-13T23:59:59.000Z

479

Z-Area Saltstone Disposal Facility groundwater monitoring report. 1996 annual report  

SciTech Connect (OSTI)

The Z-Area Saltstone Disposal Facility is located in the Separations Area, north of H and S Areas, at the Savannah River Site (SRS). The facility permanently disposes of low-level radioactive waste. The facility blends low-level radioactive salt solution with cement, slag, and flyash to form a nonhazardous cementitious waste that is pumped to aboveground disposal vaults. Z Area began these operations in June 1990. Samples from the ZBG wells at the Z-Area Saltstone Disposal Facility are analyzed for constituents required by South Carolina Department of Health and Environmental Control (SCDHEC) Industrial Solid Waste Permit {number_sign}025500-1603 (formerly IWP-217). During second quarter 1996, lead was reported above the SCDHEC-proposed groundwater monitoring standard in one well. No other constituents were reported above SCDHEC-proposed groundwater monitoring standards for final Primary Drinking Water Standards during first, second, or third quarters 1996. Antimony was detected above SRS flagging criteria during third quarter 1996. In the past, tritium has been detected sporadically in the ZBG wells at levels similar to those detected before Z Area began radioactive operations.

NONE

1996-12-01T23:59:59.000Z

480

Evaluation of the tritium content in light water reactor control and absorber rods to obtain data for the fuel cycle backend  

SciTech Connect (OSTI)

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/%.

Bleier, A.; Neeb, K.H.; Gelfort, E.; Mischke, J.

1986-08-01T23:59:59.000Z

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


481

The Development of RF Heating of Magnetically Confined Deuterium-Tritium Plasmas  

SciTech Connect (OSTI)

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