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APT/LEDA RFQ and support frame structural analysis

Description: This report documents structural analysis of the Accelerator Production of Tritium Low Energy Demonstration Accelerator (APT/LEDA) Radio Frequency Quadrupole (RFQ) accelerator structure and its associated support frame. This work was conducted for the Department of Energy in support of the APT/LEDA. Structural analysis of the RFQ was performed to quantify stress levels and deflections due to both vacuum loading and gravity loading. This analysis also verified the proposed support scheme geometry and quantified interface loads. This analysis also determined the necessary stiffness and strength requirements of the RFQ support frame verifying the conceptual design geometry and allowing specification of individual frame elements. Complete structural analysis of the frame was completed subsequently. This report details structural analysis of the RFQ assembly with regard to gravity and vacuum loads only. Thermally induced stresses from the Radio Frequency (RF) surface resistance heating were not considered.
Date: April 1, 1997
Creator: Ellis, S.
Partner: UNT Libraries Government Documents Department

Tritium experience in the Tokamak Fusion Test Reactor

Description: Tritium management is a key enabling element in fusion technology. Tritium fuel was used in 3.5 years of successful deuterium-tritium (D-T) operations in the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory. The D-T campaign enabled TFTR to explore the transport, alpha physics, and MHD stability of a reactor core. It also provided experience with tritium retention and removal that highlighted the importance of these issues in future D-T machines. In this paper, the authors summarize the tritium retention and removal experience in TFTR and its implications for future reactors.
Date: July 1, 1998
Creator: Skinner, C.H.; Blanchard, W.; Hosea, J.; Mueller, D.; Nagy, A.; Brooks, J.N. et al.
Partner: UNT Libraries Government Documents Department

Tritium Removal by CO{sub 2} Laser Heating

Description: Efficient techniques for rapid tritium removal will be necessary for ITER (International Thermonuclear Experimental Reactor) to meet its physics and engineering goals. One potential technique is transient surface heating by a scanning CO(subscript 2) or Nd:Yag laser that would release tritium without the severe engineering difficulties of bulk heating of the vessel. We have modeled the heat propagation into a surface layer and find that a multi-kW/cm(superscript2) flux with an exposure time of order 10 msec is suitable to heat a 50 micron co-deposited layer to 1,000-2,000 degrees. Improved wall conditioning may be a significant side benefit. We identify remaining issues that need to be addressed experimentally.
Date: October 1, 1997
Creator: Doyle, B.L.; Skinner, C.H.; Mueller, D.; Kugel, H. & Wampler, W.R.
Partner: UNT Libraries Government Documents Department

Tritium retention and removal on TFTR

Description: Tritium retention and removal are critical issues for the success of ITER or any DT fusion reactor. The Tokamak Fusion Test Reactor, TFTR, is the first fusion facility to afford the opportunity to study the tritium retention and removal over an extended period. In TFTR, tritium accumulates on all surfaces with line of sight to the plasma by codeposition of tritium with carbon. Measurements of both deuterium and tritium retention fractions have yielded retention between 0.2 and 0.6 of the injected fuel in the torus. Tritium has been successfully removed from TFTR by glow discharge cleaning and by air purges. The in-vessel inventory was reduced by a factor of 2, facilitating machine maintenance. In TFTR, the amount of dust recovered from the TFTR vacuum vessel has varied from several grams to a few kilograms.
Date: November 1, 1997
Creator: Mueller, D.; Blanchard, W. & Doyle, B.L.
Partner: UNT Libraries Government Documents Department

Studies of tritiated co-deposited layers in TFTR

Description: Plasma facing components in TFTR contain an important record of plasma wall interactions in reactor grade DT plasmas. Tiles, flakes, wall coupons, a stainless steel shutter and dust samples have been retrieved from the TFTR vessel for analysis. Selected samples have been baked to release tritium and assay the tritium content. The in-vessel tritium inventory is estimated to be 0.56 g and is consistent with the in-vessel tritium inventory derived from the difference between tritium fueling and tritium exhaust. The distribution of tritium on the limiter and vessel wall showed complex patterns of co-deposition. Relatively high concentrations of tritium were found at the top and bottom of the bumper limiter, as predicted by earlier BBQ modeling.
Date: June 28, 2000
Creator: Skinner, C.H.; Gentile, C.A.; Ascione, G.; Carpe, A.; Causey, R.A.; Hayashi, T. et al.
Partner: UNT Libraries Government Documents Department

Comparison of methods for separating small quantities of hydrogen isotopes from an inert gas

Description: It is frequent within tritium processing systems that a small amount of hydrogen isotopes (Q{sub 2}) must be separated from an inert gas such as He, Ar and N{sub 2}. Thus, a study of presently available technologies for effecting such a separation was performed. A base case and seven technology alternatives were identified and a simple design of each was prepared. These technologies included oxidation-adsorption-metal bed reduction, oxidation-adsorption-palladium membrane reactor, cryogenic adsorption, cryogenic trapping, cryogenic distillation, hollow fiber membranes, gettering and permeators. It was found that all but the last two methods were unattractive for recovering Q{sub 2} from N{sub 2}. Reasons for technology rejection included (1) the method unnecessarily turns the hydrogen isotopes into water, resulting in a cumbersome and more hazardous operation, (2) the method would not work without further processing, and (3) while the method would work, it would only do so in an impractical way. On the other hand, getters and permeators were found to be attractive methods for this application. Both of these methods would perform the separation in a straightforward, essentially zero-waste, single step operation. The only drawback for permeators was that limited low-partial Q{sub 2} pressure data is available. The drawbacks for getters are their susceptibility to irreversible and exothermic reaction with common species such as oxygen and water, and the lack of long-term operation of such beds. More research is envisioned for both of these methods to mature these attractive technologies.
Date: March 1, 1998
Creator: Willms, R.S.; Tuggle, D.; Birdsell, S.; Parkinson, J.; Price, B. & Lohmeir, D.
Partner: UNT Libraries Government Documents Department

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

Description: 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.
Date: April 1, 1997
Creator: Sze, D.K.; Hua, T.Q.; Abdou, M.A.; Dagher, M.A. & Waganer, L.M.
Partner: UNT Libraries Government Documents Department

Low energy demonstration accelerator technical area 53

Description: As part of the Department of Energy`s (DOE) need to maintain the capability of producing tritium in support of its historic and near-term stewardship of the nation`s nuclear weapons stockpile, the agency has recently completed a Programmatic Environmental Impact Statement for Tritium Supply and Recycling. The resulting Record of Decision (ROD) determined that over the next three years the DOE would follow a dual-track acquisition strategy that assures tritium production for the nuclear weapon stockpile in a rapid, cost effective, and safe manner. Under this strategy the DOE will further investigate and compare two options for producing tritium: (1) purchase of an existing commercial light-water reactor or irradiation services with an option to purchase the reactor for conversion to a defense facility; and (2) design, build, and test critical components of a system for accelerator production of tritium (APT). The final decision to select the primary production option will be made by the Secretary of Energy in the October 1998 time frame. The alternative not chosen as the primary production method, if feasible, would be developed as a back-up tritium supply source. This Environmental Assessment (EA) analyzes the potential environmental effects that would be expected to occur if the DOE were to design, build, and test critical prototypical components of the accelerator system for tritium production, specifically the front-end low-energy section of the accelerator, at Los Alamos National Laboratory. The Low Energy Demonstration Accelerator (LEDA) would be incrementally developed and tested in five separate stages over the next seven years. The following issues were evaluated for the proposed action: utility demands, air, human health, environmental restoration, waste management, transportation, water, threatened and endangered species, wetlands, cultural resources, and environmental justice.
Date: April 1, 1996
Partner: UNT Libraries Government Documents Department

Status of the Accelerator Production of Tritium (APT) project

Description: Tritium is a radioactive isotope of hydrogen used in all United States nuclear weapons. Because the half-life of tritium is short, 12.3 years. it must be periodically replaced. To provide a new source, the United States Department of Energy is sponsoring conceptual design and engineering development and demonstration activities for a plant that will use a high-power proton linear accelerator to produce tritium. This paper presents an overview of activities planned or underway to support that work.
Date: July 1, 1996
Creator: Browne, J.C.; Anderson, J.L. & Cappiello, M.W.
Partner: UNT Libraries Government Documents Department

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

Description: A 2-stage cold (non-tritium) PMR system was tested with the ITER mix in61 days of continuous operation. No decrease in performance was observed over the duration of the test. Decontamination factor (DF) was found to increase with decreasing inlet rate. Decontamination factors in excess of 1.4 {times} 10{sup 5} were obtained, but the exact value of the highest DF could not be determined because of analysis limitations. Results of the 61-day test were used to design a 2-stage PMR system for use in tritium testing. The PMR system was scaled up by a factor of 6 and built into a glovebox in the Tritium Systems Test Assembly (TSTA) of the Los Alamos National Laboratory. This system is approximately 1/5th of the expected full ITER scale. The ITER mix was injected into the PMR system for 31 hours, during which 4.5 g of tritium were processed. The 1st stage had DF = 200 and the 2nd stage had DF = 2.9 {times} 10{sup 6}. The overall DF = 5.8 {times} 10{sup 8}, which is greater than ITER requirements.
Date: December 1996
Creator: Birdsell, S. A.; Willms, R. S. & Wilhelm, R. C.
Partner: UNT Libraries Government Documents Department

Tritium removal by CO{sub 2} laser heating

Description: Efficient techniques for rapid tritium removal will be necessary for ITER (International Thermonuclear Experimental Reactor) to meet its physics and engineering goals. One potential technique is transient surface heating by a scanning CO{sub 2} or Nd:YAG laser that would release tritium without the severe engineering difficulties of bulk heating of the vessel. The authors have modeled the heat propagation into a surface layer and find that a multi-kW/cm{sup 2} flux with an exposure time of order 10 msec is suitable to heat a 50 micron co-deposited layer to 1,000--2,000 degrees. Improved wall conditioning may be a significant side benefit. They identify remaining issues that need to be addressed experimentally.
Date: October 1, 1997
Creator: Skinner, C.H.; Kugel, H.; Mueller, D.; Doyle, B.L. & Wampler, W.R.
Partner: UNT Libraries Government Documents Department

Tritium removal by CO{sub 2} laser heating

Description: Efficient techniques for rapid tritium removal will be necessary for ITER to meet its physics and engineering goals. One potential technique is transient surface heating by a scanning CO{sub 2} or Nd:Yag laser that would release tritium without the severe engineering difficulties of bulk heating of the vessel. The authors have modeled the heat propagation into a surface layer and find that a multi-kW/cm{sup 2} flux with an exposure time of order 10 ms is suitable to heat a 50 micron co-deposited layer to 1,000--2,000 degrees. Improved wall conditioning may be a significant side benefit. They identify remaining issues that need to be addressed experimentally.
Date: October 1, 1997
Creator: Skinner, C.H.; Kugel, H.; Mueller, D.; Doyle, B.L. & Wampler, W.R.
Partner: UNT Libraries Government Documents Department

Thermal conductivity and tritium retention in Li{sub 2}O and Li{sub 2}ZrO{sub 3}

Description: Lithium oxide (Li{sub 2}O) and lithium zirconate (Li{sub 2}ZrO{sub 3}) are promising ceramic breeder materials for fusion reactor blankets. The thermal and tritium transport databases for these materials are reviewed. Algorithms are presented for predicting both the temperature distribution and the retained tritium profile across sintered-product and pebble-bed regions. Sample design calculations are also performed to demonstrate the relative advantages of each breeder ceramic. For Li{sub 2}O, the thermal conductivity of sintered-product material has been measured over a wide range of temperatures and densities. Data are also available for the effective thermal conductivity of a pebble bed (in atmospheric helium) with 55% packing fraction for the 5-mm-diameter/75%-dense pebbles. Similar results are available for sintered-product and pebble-bed (60% packing fraction for 1.2-mm-diameter/80%-dense pebbles in atmospheric He) Li{sub 2}ZrO{sub 3}. Hall and Martin model predictions are in reasonable agreement with both sets of pebble bed data. Thus, the databases and calculational algorithms are well established for performing thermal analyses. 15 refs., 5 figs.
Date: August 1, 1997
Creator: Billone, M.C.
Partner: UNT Libraries Government Documents Department

A tritium vessel cleanup experiment in TFTR

Description: A simple tritium cleanup experiment was carried out in TFTR following the initial high power deuterium-tritium discharges in December 1993. A series of 34 ohmic and deuterium neutral beam fueled shots was used to study the removal of tritium implanted into the wall and limiters. A very large plasma was created in each discharge to ``scrub`` an area as large as possible. Beam-fueled shots at 2.5 to 7.5 MW of injected power were used to monitor tritium concentration levels in the plasma by detection of DT-neutrons. The neutron signal decreased by a factor of 4 during the experiment, remaining well above the expected T-burnup level. The amount of tritium recovered at the end of the cleanup was about 8% of the amount previously injected with high power DT discharges. The experience gained suggests that measurements of tritium inventory in the torus are very difficult to execute and require dedicated systems with overall accuracy of 1%.
Date: March 1, 1995
Creator: Caorlin, M.; Kamperschroer, J.; Owens, D.K.; Voorhees, D.; Mueller, D.; Ramsey, A.T. et al.
Partner: UNT Libraries Government Documents Department

Overview of design activities for Li/V blankets

Description: Recent fusion power plant design studies in the US have been conducted within the ARIES project. The most recent design of Li/V blankets was conducted as part of the ARIES-RS design. The ARIES-RS fusion power plant design study is based on reversed-shear (RS) physics with a Li/V (lithium breeder and vanadium structure) blanket. The reversed-shear discharge has been documented in many large tokamak experiments. The plasma in the RS mode has a high beta, low current, and low current drive requirement. Therefore, it is an attractive physics regime for a fusion power plant. The blanket system based on a Li/V has high temperature operating capability, good tritium breeding, excellent high heat flux removal capability, long structural life time, low activation, low after heat and good safety characteristics. For these reasons, the ARIES-RS reactor study selected Li/V as the reference blanket. The combination of attractive physics and attractive blanket engineering is expected to result in a superior power plant design.
Date: December 31, 1997
Creator: Sze, D.K. & Mattas, R.F.
Partner: UNT Libraries Government Documents Department

Tritium activities in the United States

Description: There have been many significant changes in the status of tritium activities in the US since the 4th Tritium Conference in October, 1991. The replacement Tritium Facility (RTF) at Savannah River Site and the Weapons Engineering Tritium Facility (WETF) at the Los Alamos National Laboratory are now operational with tritium. The Tokamak Fusion Test Reactor (TFTR) has initiated a highly successful experimental campaign studying DT plasmas, and has produced more than 10 Megawatts (MW) of fusion power in a D-T plasma. Sandia National Laboratory has ceased tritium operations at the Tritium Research Laboratory (TRL) and many of the activities previously performed there have been transferred to Los Alamos and Savannah River. The tritium laboratory at Lawrence Livermore National Laboratory has reduced the tritium inventory to <5 grams. The Tritium Systems Test Assembly (TSTA) at Los Alamos continues to be at the forefront of tritium technology and safety development for the fusion energy program.
Date: July 1, 1995
Creator: Anderson, J.L. & LaMarche, P.
Partner: UNT Libraries Government Documents Department

Investigations of the tritium recycling in TFTR using the DT neutron rate

Description: During deuterim-only neutral-beam-injected discharges, tritium from earlier deuterium-tritium discharges is released from the vessel limiters and walls to cause a deuterium-tritium neutron count rate comparable to the deuterium-deuterium neutron count rate. A measure of the tritium density in the plasma based on neutron rate measurements is defined and used to determine which parameters influence tritium influx to the plasma core. The tritium density is observed to decrease in a sequence of deuterium-only supershots and to depend on the amount of tritium injected in prior DT shots and the amount of tritium present in the limiter. A weak correlation is also observed with the plasma current, but not with beam power, hydrogen influx, carbon influx, visible bremsstrahlung, lithium pellet injection, blooms, nor disruptions.
Date: April 1, 1996
Creator: Kruger, S. E.; Callen, J. D.; Budny, R. V.; Chang, Z.; Skinner, C. H. & Strachan, J. D.
Partner: UNT Libraries Government Documents Department

Code development incorporating environmental, safety and economic aspects of fusion reactors; Annual progress report

Description: This document is a proposal to continue the authors work on the Environmental, Safety and Economic (ESE) aspects of fusion reactors under DOE contract DE-FR03-89ER52514. The grant objectives continue those from the previous grant: (1) completion of first-generation Environmental, Safety and Economic (ESE) computer modules suitable as integral components of tokamak systems codes. (2) continuation of work on special topics, in support of the above and in response to OFE requests. The proposal also highlights progress on the contract in the twelve months since April, 1992. This has included work with the ARIES and ITER design teams, work on tritium management, studies on materials activation, and calculation of radioactive inventories in fusion reactors.
Date: December 31, 1993
Creator: Fowler, T.K.; Greenspan, E. & Holdren, J.P.
Partner: UNT Libraries Government Documents Department

Argon frost continuous cryopump for fusion applications

Description: A cryopumping system based on the snail continuous cryopump concept is being developed for fusion applications under a DOE SBIR grant. The primary pump is a liquid helium cooled compound pump designed to continuously pump and fractionate deuterium/tritium and helium. The D/T pumping stage is a 500 mm bore cryocondensation pump with a nominal pumping speed of 45,000 L/s. It will be continuously regenerated by a snail regeneration by head every 12 minutes. Continuous regeneration will dramatically reduce the vulnerable tritium inventory in a fusion reactor. Operating at an inlet pressure of 1 millitorr, eight of these pumps could pump the projected D/T flow in the ITER CDA design while reducing the inventory of tritium in the pumping system from 630 to 43 grams. The helium fraction will be pumped in a compound argon frost stage. This stage will also operate continuously with a snail regeneration head. In addition the argon spray head will be enclosed inside the snail, thereby removing gaseous argon from the process chamber. Since the cryocondensation stage will intercept over 90% of the D/T/H steam, a purified stream from this stage could be directly reinjected into the plasma as gas or pellets, thereby bypassing the isotope separation system and further simplifying the fuel cycle. Experiments were undertaken in Phase I which demonstrated continuous cryosorption pumping of hydrogen on CO{sub 2} and argon frosts. The pumping system and its relevance to fusion reactor pumping will be discussed.
Date: December 1, 1993
Creator: Foster, C. A. & McCurdy, H. C.
Partner: UNT Libraries Government Documents Department

BEATRIX-II: In-situ tritium recovery data correction

Description: BEATRIX-II was an in-situ tritium recovery experiment in a fast reactor to characterize the irradiation behavior of fusion ceramic breeder materials. Correcting and compiling the in-situ tritium recovery data involved correcting the ion chamber response for the effect of sweep gas composition or amount of hydrogen in the helium sweep gas and for the buildup of background. The effect of sweep gas composition was addressed in the previous workshop. During the operation of Phase I of the experiment the backgrounds of the ion chambers were found to reach significant levels relative to the tritium recovery concentrations in the sweep gas from the specimen canisters. The measured tritium concentrations were corrected for background by comparing the tritium recovery rate during reference conditions with the predicted tritium generation rate. Background increases were found to be associated with tritium recovery peaks and elevated levels of moisture in the sweep gas. These conditions typically occurred when the hydrogen concentration in the sweep gas was increased to 0.1% after extended operation in He or He-0.01% H{sub 2}. Three examples of this increase in ionization chamber background are described. The final corrected BEATRIX-II, Phase I tritium recovery data provide a valuable resource to be used for predicting the performance of Li{sub 2}O in a fusion blanket application.
Date: September 1, 1993
Creator: Slagle, O. D.; Hollenberg, G. W.; Kurasawa, T. & Verrall, R. A.
Partner: UNT Libraries Government Documents Department