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Absolute Intensities of the Vacuum Ultraviolet Spectra in a Metal-Etch Plasma Processing Discharge

Description: In this paper we report absolute intensities of vacuum ultraviolet and near ultraviolet emission lines (4.8 eV to 18 eV ) for aluminum etching discharges in an inductively coupled plasma reactor. We report line intensities as a function of wafer type, pressure, gas mixture and rf excitation level. IrI a standard aluminum etching mixture containing C12 and BC13 almost all the light emitted at energies exceeding 8.8 eV was due to neutral atomic chlorine. Optical trapping of the WV radiation in the discharge complicates calculations of VUV fluxes to the wafer. However, we see total photon fluxes to the wailer at energies above 8.8 eV on the order of 4 x 1014 photons/cm2sec with anon- reactive wafer and 0.7 x 10 `4 photons/cm2sec with a reactive wtier. The maj ority of the radiation observed was between 8.9 and 9.3 eV. At these energies, the photons have enough energy to create electron-hole pairs in Si02, but may penetrate up to a micron into the Si02 before being absorbed. Relevance of these measurements to vacuum-W photon-induced darnage of Si02 during etching is discussed.
Date: December 9, 1998
Creator: Aragon, B.P.; Blain, M.G.; Hamilton, T.W.; Jarecki, R.L. & Woodworth, J.R.
Partner: UNT Libraries Government Documents Department

Testing of sludge coating adhesiveness on fuel elements in 105-K west basin

Description: This report summarizes the results from the first sludge adherence tests performed in the 105-K West Basin on N Reactor fuel. The outside surface of the outer fuel elements were brushed, using stainless steel wire brushes, to test the adhesiveness of various types of sludge coatings to the cladding`s surface. The majority of the sludge was removed by the wire brushes in this test but different types of sludge were more adhesive than others. Particularly, an orange rust-like sludge coating that was just slightly more adherent to the fuel`s cladding than the majority of the sludge coatings and a thick white vertical strip sludge coating that was much more difficult to remove. The test demonstrated that all of the sludge could be removed from the outer fuel elements` surfaces if the need arises.
Date: March 11, 1997
Creator: Maassen, D.P., Fluor Daniel Hanford
Partner: UNT Libraries Government Documents Department

University of Florida--US Department of Energy 1994-1995 reactor sharing program

Description: The grant support of $24,250 (1994-95?) was well used by the University of Florida as host institution to support various educational institutions in the use of UFTR Reactor. All users and uses were screened to assure the usage was for educational institutions eligible for participation in the Reactor Sharing Program; where research activities were involved, care was taken to assure the research was not funded by grants for contract funding from outside sources. Over 12 years, the program has been a key catalyst for renewing utilization of UFTR both by external users around the State of Florida and the Southeast and by various faculty members within the University of Florida. Tables provide basic information about the 1994-95 program and utilization of UFTR.
Date: June 1, 1996
Creator: Vernetson, W.G.
Partner: UNT Libraries Government Documents Department

Advanced reactors transition FY 1997 multi-year work plan WBS 7.3

Description: This document describes in detail the work to be accomplised in FY 1997 and the out-years for the Advanced Reactors Transition (WBS 7.3) under the management of the Babcock & Wilcox Hanford Company. This document also includes specific milestones and funding profiles. Based upon the Fiscal Year 1997 Multi-Year Work Plan, the Department of Energy will provide authorization to perform the work described.
Date: September 27, 1996
Creator: Hulvey, R.K.
Partner: UNT Libraries Government Documents Department

Technical Safety Requirements for the Annular Core Research Reactor Faility (ACRRF)

Description: The Technical Safety Requirements (TSR) document is prepared and issued in compliance with DOE Order 5480.22, Technical Safety Requirements. The bases for the TSR are established in the ACRRF Safety Analysis Report issued in compliance with DOE Order 5480.23, Nuclear Safety Analysis Reports. The TSR identifies the operational conditions, boundaries, and administrative controls for the safe operation of the facility.
Date: September 1, 1998
Creator: Boldt, K.R.; McCrory, F.M.; Morris, F.M. & Talley, D.G.
Partner: UNT Libraries Government Documents Department

High flux isotope reactor redesigned beryllium reflector thermal stress calculations

Description: The Beryllium reflector of the High Flux Isotope Reactor is currently redesigned for upgrading the capability of the reactor. The original design criteria are adopted in the redesign analysis. Both nuclear heating and thermal stress calculations are revised. The results show that more margin of safety have been achieved and the updated design assures more precise design estimates for the reflector thermal stress conditions. 1 ref., 2 figs., 2 tabs.
Date: June 1, 1996
Creator: Chang, S.J.
Partner: UNT Libraries Government Documents Department

Characterization of the 309 fuel examination facility

Description: This document identifies radiological, chemical and physical conditions inside the Fuel Examination Facility. It is located inside the Plutonium Recycle Test Reactor containment structure (309 Building.) The facility was a hot cell used for examination of PRTR fuel and equipment during the 1960`s. Located inside the cell is a PRTR shim rod assembly, reported are radiological conditions of the sample. The conditions were assessed as part of overall 309 Building transition.
Date: July 9, 1997
Creator: Greenhalgh, W.O. & Cornwell, B.C.
Partner: UNT Libraries Government Documents Department

Analyses of a Reinforced Concrete Containment with Liner Corrosion Damage

Description: Incidents of liner corrosion in nuclear power containment structures have been recorded. These incidents and concerns of other possible liner corrosion in containment have prompted an interest in determining g the capacity of a degraded containment. Finite element analyses of a typical pressurized water reactor (PWR) reinforced concrete containment with liner corrosion were conducted using the A13AQUS finite element code with the ANACAP-U nonlinear concrete constitutive model. The effect of liner corrosion on containment capacity was investigated. A loss of coolant accident was simulated by applying pressure and temperature changes to the structure without corrosion to determine baseline failure limits, followed by multiple analyses of the containment with corrosion at different locations and varying degrees of liner degradation. The corrosion locations were chosen at the base of the containment wall, near the equipment hatch, and at the midheight of the containment wall. Using a strain-based failure criterion the different scenarios were evaluated to prioritize their effect on containment capacity
Date: November 19, 1998
Creator: Cherry, J.L. & Smith, J.A.
Partner: UNT Libraries Government Documents Department

Potential role of the Fast Flux Test Facility and the advanced test reactor in the U.S. tritium production system

Description: The Deparunent of Energy is currently engaged in a dual-track strategy to develop an accelerator and a conunercial light water reactor (CLWR) as potential sources of tritium supply. New analysis of the production capabilities of the Fast Flux Test Facility (FFTF) at the Hanford Site argues for considering its inclusion in the tritium supply,system. The use of the FFTF (alone or together with the Advanced Test Reactor [ATR] at the Idaho National Engineering Laboratory) as an integral part of,a tritium production system would help (1) ensure supply by 2005, (2) provide additional time to resolve institutional and technical issues associated with the- dual-track strategy, and (3) reduce discounted total life-cycle`costs and near-tenn annual expenditures for accelerator-based systems. The FFRF would also provide a way to get an early start.on dispositioning surplus weapons-usable plutonium as well as provide a source of medical isotopes. Challenges Associated With the Dual-Track Strategy The Departinent`s purchase of either a commercial reactor or reactor irradiation services faces challenging institutional issues associated with converting civilian reactors to defense uses. In addition, while the technical capabilities of the individual components of the accelerator have been proven, the entire system needs to be demonstrated and scaled upward to ensure that the components work toge ther 1548 as a complete production system. These challenges create uncertainty over the ability of the du2a-track strategy to provide an assured tritium supply source by 2005. Because the earliest the accelerator could come on line is 2007, it would have to operate at maximum capacity for the first few years to regenerate the reserves lost through radioactive decay aftei 2005.
Date: October 1, 1996
Creator: Dautel, W. A.
Partner: UNT Libraries Government Documents Department

Four in. three jaw type connector suitable for vertical mounting sample for job 11

Description: As authorized by purchase order No. 11-2532 dated March 11, 1949 of the Kellex Corp., the first sample of a 4 inch three jaw type connector for Job 11 was fabricated. The design of the connector was per Crane Co. drawing DR-25126-D except vertical mounting. The materials were per Crane Co. drawing A-24491-C. As instructed in verbal conversation with Dr. D.D. Jacobus and as requested in Mr. J.J. Cuniffe`s letter of May 6, 1949, and Ingersoll-Rand Company`s 1 1/4 inch 534 impact wrench with suitable socket to fit the hexagon head of the operating screw of the 4 inch connector was procured on loan to make some preliminary tests. Engineering drawings are listed in a second report on the data base.
Date: July 8, 1949
Creator: Grubbe, A.C.
Partner: UNT Libraries Government Documents Department

IN-CELL visual examinations of K east fuel elements

Description: Nine outer fuel elements were recovered from the K East Basin and transferred to a hot cell for examination. Extensive testing planned for these elements will support the process design for the Integrated Process Strategy (IPS), with emphasis on drying and conditioning behavior. Visual examinations of the fuel elements confirmed that they are appropriate to meet testing objectives to provide design guidance for IPS processing parameters.
Date: March 6, 1997
Creator: Pitner, A.L. & Pyecha, T.D., Fluor Daniel Hanford
Partner: UNT Libraries Government Documents Department

The sharing of the Penn State Breazeale Reactor with other educational institutions

Description: The Penn State Radiation Science and Engineering Center (RSEC) integrates the Breazeale Reactor and its affiliated laboratories and facilities on the University Park Campus. Penn State has the only nuclear reactor in Pennsylvania dedicated to research and education. Its faculty have pioneered industrial and research applications of radiation and radioisotopes. In addition, the center and its affiliated faculty have access to the multidisciplinary resources and expertise available within Penn State, one of the nation`s leading research universities. The goals of the Penn State Radiation Science and Engineering Center are to: incorporate radiation science and engineering services and facilities into a cohesive infrastructure; provide state-of-the-art academic instruction and laboratory experiences; provide facilities and assistance for academic research; provide technical, engineering, and other support services to RSEC users; generate new techniques, applications, and services for researchers in diverse disciplines; and serve the needs of academia and industry through RSEC services, faculty affiliations, and facilities that are not readily available elsewhere.
Date: December 31, 1995
Partner: UNT Libraries Government Documents Department

Time delays between core power production and external detector response from Monte Carlo calculations

Description: One primary concern for design of safety systems for reactors is the time response of external detectors to changes in the core. This paper describes a way to estimate the time delay between the core power production and the external detector response using Monte Carlo calculations and suggests a technique to measure the time delay. The Monte Carlo code KENO-NR was used to determine the time delay between the core power production and the external detector response for a conceptual design of the Advanced Neutron Source (ANS) reactor. The Monte Carlo estimated time delay was determined to be about 10 ms for this conceptual design of the ANS reactor.
Date: August 1, 1996
Creator: Valentine, T. E. & Mihalczo, J. T.
Partner: UNT Libraries Government Documents Department

RLA room 20 cleanout and stabilization

Description: This engineering report documents the decontamination and stabilization of the Rupture Loop Annex located in room 20 of the 309 building`s Plutonium Recycle Test Reactor. Low level, mixed, and recyclable waste was removed from the room. Smearable contamination was removed and/or fixed in place with paint. The RLA was cleaned out and stabilized to meet the Environmental Restoration Contractor`s turnover criteria.
Date: June 1, 1997
Creator: Ham, J.E.
Partner: UNT Libraries Government Documents Department

FFTF Plant transition function analysis report

Description: The document contains the functions, function definitions, function interfaces, function interface definitions, Input Computer Automated Manufacturing Definition (IDEFO) diagrams, and function hierarchy charts that describe what needs to be performed to deactivate FFTF.
Date: September 1, 1995
Creator: Lund, D.P. & Group, FFTF Working
Partner: UNT Libraries Government Documents Department

SCALE Graphical Developments for Improved Criticality Safety Aalyses

Description: New computer graphic developments at Oak Ridge National Ridge National Laboratory (ORNL) are being used to provide visualization of criticality safety models and calculational results as well as tools for criticality safety analysis input preparation. The purpose of this paper is to present the status of current development efforts to continue to enhance the SCALE (Standardized Computer Analyses for Licensing Evaluations) computer software system. Applications for criticality safety analysis in the areas of 3-D model visualization, input preparation and execution via a graphical user interface (GUI), and two-dimensional (2-D) plotting of results are discussed.
Date: September 20, 1999
Creator: Barnett, D. L.; Bowman, S. M.; Horwedel, J. E. & Petrie, L. M.
Partner: UNT Libraries Government Documents Department