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Program status 2. quarter -- FY 1990: Fusion technology development

Description: During this period, the ARIES-I blanket design team has concentrated its efforts on preparation of the final report. For the ARIES-II blanket design, two concepts are being evaluated. They are the Li self-cooled and the helium-cooled lithium breeder designs. The scoping design of the second concept has been completed. Varian EIMAC has had two tube failures in trying to assemble the X2274 tetrodes for the tests in Japan. Despite the failures it is still possible for the tubes to be ready as scheduled. Also during this quarter, the joint US/PRC integral experiment on beryllium was completed in March and the analysis of results has begun. Finally, the final design of DIII-D Divertor Material Exposure System (DiMES) was completed. Preliminary analysis by ANL of DIII-D divertor erosion, using measured plasma conditions, predicts maximum net erosion of 50 {micro}m and maximum net deposition of 23 {micro}m. Measurement by SNL-L of the 12 tiles removed in December 1989 is still pending.
Date: May 1, 1990
Partner: UNT Libraries Government Documents Department

Bolometry for divertor characterization and control

Description: Operation of the divertor will provide one of the greatest challenges for ITER. Up to 400 MW of power is expected to be produced in the core plasma which must then be handled by plasma facing components. Power flowing across the separatrix and into the scrape-off-layer (SOL) can lead to a heat flux in the divertor of 30 MW/m{sup 2} if nothing is done to dissipate the power. This peak heat flux must be reduced to 5 MW/m{sup 2} for an acceptable engineering design. The current plan is to use impurity radiation and other atomic processes from intrinsic or injected impurities to spread out the power onto the first wall and divertor chamber walls. It is estimated that 300 MW of radiation in the divertor and SOL will be necessary to achieve this solution. Measurement of the magnitude and distribution of this radiated power with bolometry will be important for understanding and controlling the nER divertor. Present experiments have shown intense regions of radiation both in the divertor near the separatrix and in the X-point region. The task of a divertor bolometer system will be to measure the distribution and magnitude of this radiation. First, radiation measurements can be used for machine protection. Intense divertor radiation will heat plasma facing surfaces that are not in direct view of temperature monitors. Measurement of the radiation distribution will provide information about the power flux to these components. Secondly, a bolometer diagnostic is a basic tool for divertor characterization and understanding. Radiation measurements are important for power accounting, as a cross check for other power diagnostics, and gross characterisation of the plasma behavior. A divertor bolometer system can provide a 2-D measurement of the radiation profile for comparison with theory and modeling. Finally a bolometer system can provide realtime signals for control of ...
Date: October 1, 1995
Creator: Leonard, A.W.; Goetz, J.; Fuchs, C.; Marashek, M.; Mast, F. & Reichle, R.
Partner: UNT Libraries Government Documents Department

DIII-D tokamak long range plan. Revision 3

Description: The DIII-D Tokamak Long Range Plan for controlled thermonuclear magnetic fusion research will be carried out with broad national and international participation. The plan covers: (1) operation of the DIII-D tokamak to conduct research experiments to address needs of the US Magnetic Fusion Program; (2) facility modifications to allow these new experiments to be conducted; and (3) collaborations with other laboratories to integrate DIII-D research into the national and international fusion programs. The period covered by this plan is 1 November 19983 through 31 October 1998.
Date: August 1, 1992
Partner: UNT Libraries Government Documents Department

Recent developments on the high power ECH installation at the DIII-D tokamak

Description: The 110 GHz gyrotron installation on the DIII-D tokamak has been upgraded to three tubes in the megawatt class with plans for further upgrades. The latest addition uses a diamond output window. The report describes the installation, plans, and experimental results to date.
Date: September 1, 1998
Creator: Lohr, J.; Ponce, D.; Callis, R.W.; Doane, J.L.; Ikezi, H. & Moeller, C.P.
Partner: UNT Libraries Government Documents Department

Vanadium alloys for the radiative divertor program of DIII-D

Description: Vanadium alloys provide an attractive solution for fusion power plants as they exhibit a potential for low environmental impact due to low level of activation from neutron fluence and a relatively short half-life. They also have attractive material properties for use in a reactor. General Atomics along with Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL), has developed a plan to utilize vanadium alloys as part of the Radiative Divertor Project (RDP) modification for the DIII-D tokamak. The goal for using vanadium alloys is to provide a meaningful step towards developing advanced materials for fusion power applications by demonstrating the in-service behavior of a vanadium alloy (V-4Cr-4Ti) in a tokamak in conjunction with developing essential fabrication technology for the manufacture of full-scale vanadium alloy components. A phased approach towards utilizing vanadium in DIII-D is being used starting with small coupons and samples, advancing to a small component, and finally a portion of the new double-null, slotted divertor will be fabricated from vanadium alloy product forms. A major portion of the program is research and development to support fabrication and resolve key issues related to environmental effects.
Date: October 1, 1995
Creator: Smith, J.P.; Johnson, W.R.; Stambaugh, R.D.; Trester, P.W.; Smith, D. & Bloom, E.
Partner: UNT Libraries Government Documents Department

APEX and ALPS, high power density technology programs in the U.S.

Description: In fiscal year (FY) 1998 two new fusion technology programs were initiated in the US, with the goal of making marked progress in the scientific understanding of technologies and materials required to withstand high plasma heat flux and neutron wall loads. APEX is exploring new and revolutionary concepts that can provide the capability to extract heat efficiently from a system with high neutron and surface heat loads while satisfying all the fusion power technology requirements and achieving maximum reliability, maintainability, safety, and environmental acceptability. ALPS program is evaluating advanced concepts including liquid surface limiters and divertors on the basis of such factors as their compatibility with fusion plasma, high power density handling capabilities, engineering feasibility, lifetime, safety and R and D requirements. The APEX and ALPS are three-year programs to specify requirements and evaluate criteria for revolutionary approaches in first wall, blanket and high heat flux component applications. Conceptual design and analysis of candidate concepts are being performed with the goal of selecting the most promising first wall, blanket and high heat flux component designs that will provide the technical basis for the initiation of a significant R and D effort beginning in FY2001. These programs are also considering opportunities for international collaborations.
Date: February 1, 1999
Creator: Wong, C.; Berk, S.; Abdou, M. & Mattas, R.
Partner: UNT Libraries Government Documents Department

Direct measurement of divertor exhaust neo enrichment in DIII-D

Description: We report first direct measurements of divertor exhaust gas impurity enrichment, {eta}{sub exh}=(exhaust impurity concentration){divided_by}(core impurity concentration), for both unpumped and D{sub 2} puff-with-divertor-pump conditions. The experiment was performed with neutral beam heated, ELMing H-mode, single-null diverted deuterium plasmas with matched core and exhaust parameters in the DIII-D tokamak. Neon gas impurity was puffed into the divertor. Neon density was measured in the exhaust by a specially modified Penning gauge and in the core by absolute charge exchange recombination spectroscopy. Neon particle accounting indicates that much of the puffed neon entered a temporary unmeasured reservoir, inferred to be the graphite divertor target, which makes direct measurements necessary to calculate divertor enrichments. D{sub 2} puff into the SOL (scrape-off layer) with pumping increased {eta}{sub exh} threefold over either unpumped conditions or D{sub 2} puff directly into the divertor with pumping. These results show that SOL flow plays an important role in divertor exhaust impurity enrichment.
Date: June 1, 1996
Creator: Schaffer, M.J.; Wade, M.R.; Maingi, R.; Monier-Garbet, P.; West, W.P.; Whyte, D.G. et al.
Partner: UNT Libraries Government Documents Department

ECH mirror interface tank for 110 GHz, 1 MW gyrotron

Description: A 1 MW, 110 GHz gyrotron is to be installed at General Optical Atomics in 1995. A Mirror Optics Unit (MOU) has been Unit designed and built to connect to the existing 110 GHz transmission line system. The unit reduces and directs a 145 mm diameter beam from the gyrotron to a 19 mm diameter beam which is then injected into a 31.8 mm diameter corrugated waveguide of the transmission line system. The unit operates under vacuum and is able to absorb beam spray from the gyrotron. The tank also contains various diagnostics equipment to protect the gyrotron and to determine the amount of energy loss in the tank, and at the window of the gyrotron output. This paper discusses further the design parameters, assembly and installation of the unit in the transmission line system.
Date: October 1, 1995
Creator: O`Neil, R.C.; Callis, R.W.; Cary, W.P.; Doane, J.L.; Gallix, R.; Hodapp, T.R. et al.
Partner: UNT Libraries Government Documents Department

Fusion technology development annual report, October 1, 1995--September 30, 1996

Description: In FY96, the General Atomics (GA) Fusion Group made significant contributions to the technology needs of the magnetic fusion program. The work is reported in the following sections on Fusion Power Plant Design Studies (Section 2), Plasma Interactive Materials (Section 3), SiC/SiC Composite Material Development (Section 4), Magnetic Diagnostic Probes (Section 5) and RF Technology (Section 6). Meetings attended and publications are listed in their respective sections. The overall objective of GA`s fusion technology research is to develop the technologies necessary for fusion to move successfully from present-day physics experiments to ITER and other next-generation fusion experiments, and ultimately to fusion power plants. To achieve this overall objective, the authors carry out fusion systems design studies to evaluate the technologies needed for next-step experiments and power plants, and they conduct research to develop basic knowledge about these technologies, including plasma technologies, fusion nuclear technologies, and fusion materials. They continue to be committed to the development of fusion power and its commercialization by US industry.
Date: March 1, 1997
Partner: UNT Libraries Government Documents Department

Improved operation of the Michelson interferometer ECE diagnostic on DIII-D

Description: The measurement of accurate temperature profiles is critical for transport analysis and equilibrium reconstruction in the DIII-D tokamak. Recent refinements in the Michelson interferometer diagnostic have produced more precise electron temperature measurements from electron cyclotron emission and made them available for a wider range of discharge conditions. Replacement of a lens-relay with a low-loss corrugated waveguide transmission system resulted in an increase in throughput of 6 dB and reduction of calibration error to around 5%. The waveguide exhibits a small polarization scrambling fraction of 0.05 at the quarter wavelength frequency and very stable transmission characteristics over time. Further reduction in error has been realized through special signal processing of the calibration and plasma interferograms.
Date: May 1, 1996
Creator: Austin, M.E.; Ellis, R.F.; Doane, J.L. & James, R.A.
Partner: UNT Libraries Government Documents Department

Visible spectroscopy in the DIII-D divertor

Description: Spectroscopy measurements in the DIII-D divertor have been carried out with a survey spectrometer which provides simultaneous registration of the visible spectrum over the region 400--900 nm with a resolution of 0.2 nm. Broad spectral coverage is achieved through use of a fiberoptic transformer assembly to map the curved focal plane of a fast (f/3) Rowland-circle spectrograph into a rastered format on the rectangular sensor area of a two-dimensional CCD camera. Vertical grouping of pixels during CCD readout integrates the signal intensity over the height of each spectral segment in the rastered image, minimizing readout time. For the full visible spectrum, readout time is 50 ms. Faster response time (< 10 ms) may be obtained by selecting for readout just a small number of the twenty spectral segments in the image on the CCD. Simultaneous recording of low charge states of carbon, oxygen and injected impurities has yielded information about gas recycling and impurity behavior at the divertor strike points. Transport of lithium to the divertor region during lithium pellet injection has been studied, as well as cumulative deposition of lithium on the divertor targets from pellet injection over many successive discharges.
Date: June 1, 1996
Creator: Brooks, N.H.; Fehling, D.; Hillis, D.L.; Klepper, C.C.; Naumenko, N.; Tugarinov, S. et al.
Partner: UNT Libraries Government Documents Department

E-SMART system for in-situ detection of environmental contaminants. Quarterly technical progress report, July--September 1997

Description: General Atomics (GA) leads a team of industrial, academic, and government organizations in the development of the Environmental Systems Management, Analysis and Reporting neTwork (E-SMART) for the Defense Advanced Research Project Agency (DARPA), by way of this Technology Reinvestment Project (TRP). E-SMART defines a standard by which networks of smart sensing, sampling, and control devices can interoperate. E-SMART{reg_sign} is intended to be an open standard, available to any equipment manufacturer. The user will be provided a standard platform on which a site-specific monitoring plan can be implemented using sensors and actuators from various manufacturers and upgraded as new monitoring devices become commercially available. This project will further develop and advance the E-SMART standardized network protocol to include new sensors, sampling systems, and graphical user interfaces.
Date: December 1, 1997
Partner: UNT Libraries Government Documents Department

E-SMART system for in-situ detection of environmental contaminants. Quarterly technical progress report, April--June 1997

Description: General Atomics (GA) leads a team of industrial, academic, and government organizations in the development of the Environmental Systems Management, Analysis and Reporting neTwork (E-SMART) for the Defense Advanced Research Project Agency (DARPA), by way of this Technology Reinvestment Project (TRP). E-SMART defines a standard by which networks of smart sensing, sampling, and control devices can interoperate. E-SMART is intended to be an open standard, available to any equipment manufacturer. The user will be provided a standard platform on which a site-specific monitoring plan can be implemented using sensors and actuators from various manufacturers and upgraded as new monitoring devices become commercially available. This project will further develop and advance the E-SMART standardized network protocol to include new sensors, sampling systems, and graphical user interfaces.
Date: August 1, 1997
Partner: UNT Libraries Government Documents Department

Nondimensional transport experiments on DIII-D and projections to an ignition tokamak

Description: The concept of nondimensional scaling of transport makes it possible to determine the required size for an ignition device based upon data from a single machine and illuminates the underlying physics of anomalous transport. The scaling of cross-field heat transport with the relative gyroradius {rho}*, the gyroradius normalized to the plasma minor radius, is of particular interest since {rho}* is the only nondimensional parameter which will vary significantly between present day machines and an ignition device. These nondimensional scaling experiments are based upon theoretical considerations which indicate that the thermal heat diffusivity can be written in the form {chi} = {chi}{sub B}{rho}*{sup x{sub {rho}}} F({beta}, v*, q, R/a, {kappa}, T{sub e}/T{sub i},...), where {chi}{sub B} = cT/eB. As explained elsewhere, x{sub {rho}} = 1 is called gyro-Bohm scaling, x{sub {rho}} is Bohm scaling, x{sub {rho}} = {minus}1/2 is Goldston scaling, and x{sub {rho}} = {minus}1 is stochastic scaling. The DIII-D results reported in this paper cover three important aspects of nondimensional scaling experiments: the testing of the underlying assumption of the nondimensional scaling approach, the determination of the {rho}* scaling of heat transport for various plasma regimes, and the extrapolation of the energy confinement time to future ignition devices.
Date: July 1, 1996
Creator: Petty, C.C.; Luce, T.C.; Balet, B.; Christiansen, J.P. & Cordey, J.G.
Partner: UNT Libraries Government Documents Department

E-SMART system for in-situ detection of environmental contaminants. Quarterly technical progress report, October--December 1997

Description: General Atomics (GA) leads a team of industrial, academic, and government organizations in the development of the Environmental Systems Management, Analysis and Reporting neTwork (E-SMART) for the Defense Advanced Research Project Agency (DARPA), by way of this Technology Reinvestment Project (TRP). E-SMART defines a standard by which networks of smart sensing, sampling, and control devices can interoperate. E-SMART{reg_sign} is intended to be an open standard, available to any equipment manufacturer. The user will be provided a standard platform on which a site-specific monitoring plan can be implemented using sensors and actuators from various manufacturers and upgraded as new monitoring devices become commercially available. This project will further develop and advance the E-SMART standardized network protocol to include new sensors, sampling systems, and graphical user interfaces.
Date: March 30, 1998
Partner: UNT Libraries Government Documents Department

MULTI-MODE ERROR FIELD CORRECTION ON THE DIII-D TOKAMAK

Description: OAK A271 MULTI-MODE ERROR FIELD CORRECTION ON THE DIII-D TOKAMAK. Error field optimization on DIII-D tokamak plasma discharges has routinely been done for the last ten years with the use of the external ''n = 1 coil'' or the ''C-coil''. The optimum level of correction coil current is determined by the ability to avoid the locked mode instability and access previously unstable parameter space at low densities. The locked mode typically has toroidal and poloidal mode numbers n = 1 and m = 2, respectively, and it is this component that initially determined the correction coil current and phase. Realization of the importance of nearby n = 1 mode components m = 1 and m = 3 has led to a revision of the error field correction algorithm. Viscous and toroidal mode coupling effects suggested the need for additional terms in the expression for the radial ''penetration'' field B{sub pen} that can induce a locked mode. To incorporate these effects, the low density locked mode threshold database was expanded. A database of discharges at various toroidal fields, plasma currents, and safety factors was supplement4ed with data from an experiment in which the fields of the n = 1 coil and C-coil were combined, allowing the poloidal mode spectrum of the error field to be varied. A multivariate regression analysis of this new low density locked mode database was done to determine the low density locked mode threshold scaling relationship n{sub e} {proportional_to} B{sub T}{sup -0.01} q{sub 95}{sup -0.79} B{sub pen} and the coefficients of the poloidal mode components in the expression for B{sub pen}. Improved plasma performance is achieved by optimizing B{sub pen} by varying the applied correction coil currents.
Date: October 1, 2002
Creator: SCOVILLE, JT & LAHAYE, RJ
Partner: UNT Libraries Government Documents Department

CHANGES IN PARTICLE PUMPING DUE TO VARIATION IN MAGNETIC BALANCE NEAR DOUBLE-NULL IN DIII-D

Description: OAK-B135 The authors report on a recent experiment examining how changes in the divertor magnetic balance affect the rate that particles can be pumped at the divertor targets. They find that both the edge density of the core plasma and divertor recycling play important roles in properly interpreting this pumping result. Previous studies on DIII-D have identified several important differences between double-null (DN) and single-null (SN) divertor operation. Small variations in the magnetic balance near-DN have large effects on both the power- and particle loadings at the divertor targets. These most likely result from an interplay between the plasma geometry and ion particle drifts, e.g., ''B x {del}B'' and ''E x B'' drifts. Other studies have shown that changes in magnetic balance affect the core plasma and where ELMs strike the vessel. In this paper, they examine how variations in the magnetic balance impact the rate at which particles are removed from the core plasma via pumping.
Date: July 1, 2003
Creator: PETRIE,TW; WATKINS,JG; ALLEN,SL; BROOKS,NH; FENSTERMACHER,ME; FERRON,JR et al.
Partner: UNT Libraries Government Documents Department

TOKAMAK EQUILIBRIA WITH CENTRAL CURRENT HOLES AND NEGATIVE CURRENT DRIVE

Description: OAK B202 TOKAMAK EQUILIBRIA WITH CENTRAL CURRENT HOLES AND NEGATIVE CURRENT DRIVE. Several tokamak experiments have reported the development of a central region with vanishing currents (the current hole). Straightforward application of results from the work of Greene, Johnson and Weimer [Phys. Fluids, 3, 67 (1971)] on tokamak equilibrium to these plasmas leads to apparent singularities in several physical quantities including the Shafranov shift and casts doubts on the existence of this type of equilibria. In this paper, the above quoted equilibrium theory is re-examined and extended to include equilibria with a current hole. It is shown that singularities can be circumvented and that equilibria with a central current hole do satisfy the magnetohydrodynamic equilibrium condition with regular behavior for all the physical quantities and do not lead to infinitely large Shafranov shifts. Isolated equilibria with negative current in the central region could exist. But equilibria with negative currents in general do not have neighboring equilibria and thus cannot have experimental realization, i.e. no negative currents can be driven in the central region.
Date: June 1, 2002
Creator: CHU, M.S. & PARKS, P.B.
Partner: UNT Libraries Government Documents Department

DIRECT ENERGY CONVERSION FISSION REACTOR FOR THE PERIOD JULY 1, 2002 THROUGH SEPTEMBER 30, 2002

Description: Direct energy conversion is the only potential means for producing electrical energy from a fission reactor without the Carnot efficiency limitations. This project was undertaken by Sandia National Laboratories, Los Alamos National Laboratories, The University of Florida, Texas A&amp;M University and General Atomics to explore the possibilities of direct energy conversion. Other means of producing electrical energy from a fission reactor, without any moving parts, are also within the statement of proposed work. This report documents the efforts of General Atomics. Sandia National Laboratories, the lead laboratory, provides overall project reporting and documentation.
Date: September 30, 2002
Creator: BROWN, L.C.
Partner: UNT Libraries Government Documents Department

THE ROLE OF NEUTRALS IN H-MODE PEDESTAL FORMATION

Description: An analytic model, derived from coupled continuity equations for the electron and neutral deuterium densities, is consistent with many features of edge electron density profiles in the DIII-D tokamak. For an assumed constant particle diffusion coefficient, the model shows that particle transport and neutral fueling produce electron and neutral density profiles that have the same characteristic scale lengths at the plasma edge. For systematic variations of density in H-mode discharges, the model predicts that the width of the electron density transport barrier decreases and the maximum gradient increases, as observed in the experiments. The widths computed from the model agree quantitatively with the experimental widths for conditions in which the model is valid. These results support models of transport barrier formation in which the H-mode particle barrier is driven by the edge particle flux and the width of the barrier is approximately the neutral penetration length.
Date: November 1, 2001
Creator: GROEBNER, R.J.; MAHDAVI, M.A.; LEONARD, A.W.; OSBORNE, T.H.; PORTER, G.D.; COLCHIN, R.J. et al.
Partner: UNT Libraries Government Documents Department

METHANE PENTRATION IN DIII-D ELMing H-MODE PLASMAS

Description: Carbon penetration into the core plasma during midplane and divertor methane puffing has been measured for DIII-D ELMing H-mode plasmas. The methane puffs are adjusted to a measurable signal, but global plasma parameters are only weakly affected (line average density, &lt;n{sub e}&gt; increases by &lt; 10%, energy confinement time, {tau}{sub E} drops by &lt; 10%). The total carbon content is derived from C{sup +6} density profiles in the core measured as a function of time using charge exchange recombination spectroscopy. The methane penetration factor is defined as the difference in the core content with the puff on and puff off, divided by the carbon confinement time and the methane puffing rate. In ELMing H-mode discharges with ion {del}B drift direction into the X-point, increasing the line averaged density from 5 to 8 x 10{sup 19} m{sup -3} dropped the penetration factor from 6.6% to 4.6% for main chamber puffing. The penetration factor for divertor puffing was below the detection limit (&lt;1%). Changing the ion {del}B drift direction to away from the X-point decreased the penetration factor by more than a factor of five for main chamber puffing.
Date: June 1, 2002
Creator: WEST, W.P.; LASNIER, C.J.; WHYTE, D.G.; ISLER, R.C.; EVANS, T.E.; JACKSON, G.L. et al.
Partner: UNT Libraries Government Documents Department

EXPLICT CALULATIONS OF HOMOCLINIC TANGLES SURROUNDING MAGNETIC ISLANDS IN TOKAMAKS

Description: We present explicit calculations of the complicated geometric objects known as homoclinic tangles that surround magnetic islands in the Poincare mapping of a tokamak's magnetic field. These tangles are shown to exist generically in the magnetic field of all toroidal confinement systems. The geometry of these tangles provides an explanation for the stochasticity known to occur near the X-points of the Poincare mapping. Furthermore, the intersection of homoclinic tangles from different resonances provides an explicit mechanism for the non-diffusive transport of magnetic field lines between these resonance layers.
Date: June 1, 2002
Creator: ROEDER, R.K.W.; RAPOPORT, B.I. & EVANS, T.E.
Partner: UNT Libraries Government Documents Department

MODELING OF STOCHASTIC MAGNETIC FLUX LOSS FROM THE EDGE OF A POOIDALLY DIVERTED TOKAMAK

Description: OAK A271 MODELING OF STOCHASTIC MAGNETIC FLUX LOSS FROM THE EDGE OF A POOIDALLY DIVERTED TOKAMAK. A field line integration code is used to study the loss of edge poloidal magnetic flux due to stochastic magnetic fields produced by an error field correction coil (C-coil) in DIII-D for various plasma shapes, coil currents and edge magnetic shear profiles. The authors find that the boundary of a diverted tokamak is more sensitive to stochastic flux loss than a nondiverted tokamak. The C-coil has been used to produce a stochastic layer in an ohmic diverted discharge with characteristics similar to those seen in stochastic boundary experiments in circular limiter ohmic plasmas, including: (1) an overall increase in recycling, (2) a broadening of the recycling profile at the divertor, and (3) a flattening of the boundary profiles over the extent of the stochastic layer predicted by the field line integration code. Profile flattening consistent with field line integration results is also seen in some high performance discharges with edge transport barriers. The prediction of a significant edge stochastic layer even in discharges with high performance and edge radial transport barriers indicates that either the self-consistent plasma response heals the stochastic layer or that edge stochastic layers are compatible with edge radial transport barriers.
Date: June 1, 2002
Creator: EVANS, TE,; MOYER, RA & MONAT, P
Partner: UNT Libraries Government Documents Department