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Corrosion control of carbon steel radioactive-liquid storage tanks

Description: As the West Valley Demonstration Project (WVDP) continues vitrification operation and begins decontamination activities, it is vital to continue to maintain the integrity of the high-level waste tanks and prevent further corrosion that may disrupt the operation. This report describes the current operational status and some corrosion concerns with corresponding control measure recommendations. 14 refs., 5 figs., 6 tabs.
Date: May 1, 1997
Creator: Chang, Ji Young
Partner: UNT Libraries Government Documents Department

Wire brush fastening device

Description: A fastening device is provided which is a variation on the conventional nut and bolt. The bolt has a longitudinal axis and threading helically affixed thereon along the longitudinal axis. A nut having a bore extending therethrough is provided. The bore of the nut has a greater diameter than the diameter of the bolt so the bolt can extend through the bore. An array of wire bristles are affixed within the bore so as to form a brush. The wire bristles extend inwardly from the bore and are constructed and arranged of the correct size, length and stiffness to guide the bolt within the bore and to restrain the bolt within the bore as required. A variety of applications of the wire brush nut are disclosed, including a bolt capture device and a test rig apparatus.
Date: August 31, 1993
Creator: Meigs, R.A.
Partner: UNT Libraries Government Documents Department

Variable depth core sampler

Description: This invention relates to a sampling means, more particularly to a device to sample hard surfaces at varying depths. Often it is desirable to take samples of a hard surface wherein the samples are of the same diameter but of varying depths. Current practice requires that a full top-to-bottom sample of the material be taken, using a hole saw, and boring a hole from one end of the material to the other. The sample thus taken is removed from the hole saw and the middle of said sample is then subjected to further investigation. This paper describes a variable depth core sampler comprimising a circular hole saw member, having longitudinal sections that collapse to form a point and capture a sample, and a second saw member residing inside the first hole saw member to support the longitudinal sections of the first member and prevent them from collapsing to form a point. The second hole saw member may be raised and lowered inside the the first hole saw member.
Date: December 31, 1994
Creator: Bourgeois, P. M. & Reger, R. J.
Partner: UNT Libraries Government Documents Department

Integrated radwaste treatment system lessons learned from 2{1/2} years of operation

Description: The Integrated Radwaste Treatment System (IRTS) at the West Valley Demonstration Project (WVDP) is a pretreatment scheme to reduce the amount of salts in the high-level radioactive waste (vitrification) stream. Following removal of cesium-137 (Cs-137) by ion-exchange in the Supernatant Treatment System (STS), the radioactive waste liquid is volume-reduced by evaporation. Trace amounts of Cs-137 in the resulting distillate are removed by ion-exchange, then the distillate is discharged to the existing plant water treatment system. The concentrated product, 37 to 41 percent solids by weight, is encapsulated in cement producing a stable, low-level waste form. The Integrated Radwaste Treatment System (IRTS) operated in this mode from May 1988 through November 1990, decontaminating 450,000 gallons of high-level waste liquid; evaporating and encapsulating the resulting concentrates into 10,393 71-gallon square drums. A number of process changes and variations from the original operating plan were required to increase the system flow rate and minimize waste volumes. This report provides a summary of work performed to operate the IRTS, including system descriptions, process highlights, and lessons learned.
Date: May 1, 1997
Creator: Baker, M.N. & Fussner, R.J.
Partner: UNT Libraries Government Documents Department

High-level waste canister storage final design, installation, and testing. Topical report

Description: This report is a description of the West Valley Demonstration Project`s radioactive waste storage facility, the Chemical Process Cell (CPC). This facility is currently being used to temporarily store vitrified waste in stainless steel canisters. These canisters are stacked two-high in a seismically designed rack system within the cell. Approximately 300 canisters will be produced during the Project`s vitrification campaign which began in June 1996. Following the completion of waste vitrification and solidification, these canisters will be transferred via rail or truck to a federal repository (when available) for permanent storage. All operations in the CPC are conducted remotely using various handling systems and equipment. Areas adjacent to or surrounding the cell provide capabilities for viewing, ventilation, and equipment/component access.
Date: April 1, 1998
Creator: Connors, B.J.; Meigs, R.A.; Pezzimenti, D.M. & Vlad, P.M.
Partner: UNT Libraries Government Documents Department

West Valley Demonstration Project site environmental report for calendar year 1996

Description: The West Valley Demonstration Project (WVDP), the site of a US Department of Energy environmental cleanup activity operated by West Valley Nuclear Services Co., Inc., (WVNS), is in the process of solidifying liquid high-level radioactive waste remaining at the site after commercial nuclear fuel reprocessing was discontinued. The Project is located in Western New York State, about 30 miles south of Buffalo, within the New York State-owned Western New York Nuclear Service Center (WNYNSC). This report represents a single, comprehensive source of off-site and on-site environmental monitoring data collected during 1996 by environmental monitoring personnel. The environmental monitoring program and results are discussed in the body of this report. The monitoring data are presented in the appendices. Appendix A is a summary of the site environmental monitoring schedule. Appendix B lists the environmental permits and regulations pertaining to the WVDP. Appendices C through F contain summaries of data obtained during 1996 and are intended for those interested in more detail than is provided in the main body of the report.
Date: June 1, 1997
Partner: UNT Libraries Government Documents Department

Design of equipment used for high-level waste vitrification at the West Valley Demonstration Project

Description: The equipment as designed, started, and operated for high-level radioactive waste vitrification at the West Valley Demonstration Project in western New York State is described. Equipment for the processes of melter feed make-up, vitrification, canister handling, and off-gas treatment are included. For each item of equipment the functional requirements, process description, and hardware descriptions are presented.
Date: June 1, 1997
Creator: Vance, R.F.; Brill, B.A. & Carl, D.E.
Partner: UNT Libraries Government Documents Department

Development of analytical cell support for vitrification at the West Valley Demonstration Project. Topical report

Description: Analytical and Process Chemistry (A&PC) support is essential to the high-level waste vitrification campaign at the West Valley Demonstration Project (WVDP). A&PC characterizes the waste, providing information necessary to formulate the recipe for the target radioactive glass product. High-level waste (HLW) samples are prepared and analyzed in the analytical cells (ACs) and Sample Storage Cell (SSC) on the third floor of the main plant. The high levels of radioactivity in the samples require handling them in the shielded cells with remote manipulators. The analytical hot cells and third floor laboratories were refurbished to ensure optimal uninterrupted operation during the vitrification campaign. New and modified instrumentation, tools, sample preparation and analysis techniques, and equipment and training were required for A&PC to support vitrification. Analytical Cell Mockup Units (ACMUs) were designed to facilitate method development, scientist and technician training, and planning for analytical process flow. The ACMUs were fabricated and installed to simulate the analytical cell environment and dimensions. New techniques, equipment, and tools could be evaluated m in the ACMUs without the consequences of generating or handling radioactive waste. Tools were fabricated, handling and disposal of wastes was addressed, and spatial arrangements for equipment were refined. As a result of the work at the ACMUs the remote preparation and analysis methods and the equipment and tools were ready for installation into the ACs and SSC m in July 1995. Before use m in the hot cells, all remote methods had been validated and four to eight technicians were trained on each. Fine tuning of the procedures has been ongoing at the ACs based on input from A&PC technicians. Working at the ACs presents greater challenges than had development at the ACMUs. The ACMU work and further refinements m in the ACs have resulted m in a reduction m in analysis turnaround ...
Date: December 1, 1997
Creator: Barber, F.H.; Borek, T.T. & Christopher, J.Z.
Partner: UNT Libraries Government Documents Department

THOREX processing and zeolite transfer for high-level waste stream processing blending

Description: The West Valley Demonstration Project (WVDP) completed the pretreatment of the high-level radioactive waste (HLW) prior to the start of waste vitrification. The HLW originated form the two million liters of plutonium/uranium extraction (PUREX) and thorium extraction (THOREX) wastes remaining from Nuclear Fuel Services` (NFS) commercial nuclear fuel reprocessing operations at the Western New York Nuclear Service Center (WNYNSC) from 1966 to 1972. The pretreatment process removed cesium as well as other radionuclides from the liquid wastes and captured these radioactive materials onto silica-based molecular sieves (zeolites). The decontaminated salt solutions were volume-reduced and then mixed with portland cement and other admixtures. Nineteen thousand eight hundred and seventy-seven 270-liter square drums were filled with the cement-wastes produced from the pretreatment process. These drums are being stored in a shielded facility on the site until their final disposition is determined. Over 6.4 million liters of liquid HLW were processed through the pretreatment system. PUREX supernatant was processed first, followed by two PUREX sludge wash solutions. A third wash of PUREX/THOREX sludge was then processed after the neutralized THOREX waste was mixed with the PUREX waste. Approximately 6.6 million curies of radioactive cesium-137 (Cs-137) in the HLW liquid were removed and retained on 65,300 kg of zeolites. With pretreatment complete, the zeolite material has been mobilized, size-reduced (ground), and blended with the PUREX and THOREX sludges in a single feed tank that will supply the HLW slurry to the Vitrification Facility.
Date: July 1, 1997
Creator: Kelly, S. Jr. & Meess, D.C.
Partner: UNT Libraries Government Documents Department

Off gas film cooler cleaner

Description: An apparatus is described for cleaning depositions of particulate matter from the inside of tubular piping while the piping is in use. The apparatus is remotely controlled in order to operate in hazardous environments. A housing containing brush and shaft assemblies is mounted on top of the tubular piping. Pneumatic cylinders provide linear motion. A roller nut bearing provides rotary motion. The combined motion causes the brush assembly to rotate as it travels along the tube dislodging particulate matter. The main application for this invention is to clean the off gas cooler of a radioactive waste vitrification unit.
Date: December 31, 1995
Creator: Dhingra, H.S.; Koch, W.C. & Burns, D.C.
Partner: UNT Libraries Government Documents Department

Vitrification Facility integrated system performance testing report

Description: This report provides a summary of component and system performance testing associated with the Vitrification Facility (VF) following construction turnover. The VF at the West Valley Demonstration Project (WVDP) was designed to convert stored radioactive waste into a stable glass form for eventual disposal in a federal repository. Following an initial Functional and Checkout Testing of Systems (FACTS) Program and subsequent conversion of test stand equipment into the final VF, a testing program was executed to demonstrate successful performance of the components, subsystems, and systems that make up the vitrification process. Systems were started up and brought on line as construction was completed, until integrated system operation could be demonstrated to produce borosilicate glass using nonradioactive waste simulant. Integrated system testing and operation culminated with a successful Operational Readiness Review (ORR) and Department of Energy (DOE) approval to initiate vitrification of high-level waste (HLW) on June 19, 1996. Performance and integrated operational test runs conducted during the test program provided a means for critical examination, observation, and evaluation of the vitrification system. Test data taken for each Test Instruction Procedure (TIP) was used to evaluate component performance against system design and acceptance criteria, while test observations were used to correct, modify, or improve system operation. This process was critical in establishing operating conditions for the entire vitrification process.
Date: May 1, 1997
Creator: Elliott, D.
Partner: UNT Libraries Government Documents Department

Cement encapsulation of low-level waste liquids. Final report

Description: Pretreatment of liquid high-level radioactive waste at the West Valley Demonstration Project (WVDP) was essential to ensuring the success of high-level waste (HLW) vitrification. By chemically separating the HLW from liquid waste, it was possible to achieve a significant reduction in the volume of HLW to be vitrified. In addition, pretreatment made it possible to remove sulfates, which posed several processing problems, from the HLW before vitrification took place.
Date: January 1, 1999
Creator: Baker, M.N. & Houston, H.M.
Partner: UNT Libraries Government Documents Department

Operating experience during high-level waste vitrification at the West Valley Demonstration Project

Description: This report provides a summary of operational experiences, component and system performance, and lessons learned associated with the operation of the Vitrification Facility (VF) at the West Valley Demonstration Project (WVDP). The VF was designed to convert stored high-level radioactive waste (HLW) into a stable waste form (borosilicate glass) suitable for disposal in a federal repository. Following successful completion on nonradioactive test, HLW processing began in July 1995. Completion of Phase 1 of HLW processing was reached on 10 June 1998 and represented the processing of 9.32 million curies of cesium-137 (Cs-137) and strontium-90 (Sr-90) to fill 211 canisters with over 436,000 kilograms of glass. With approximately 85% of the total estimated curie content removed from underground waste storage tanks during Phase 1, subsequent operations will focus on removal of tank heel wastes.
Date: January 1, 1999
Creator: Valenti, P.J. & Elliott, D.I.
Partner: UNT Libraries Government Documents Department

Radiation safety at the West Valley Demonstration Project

Description: This is a report on the Radiation Safety Program at the West Valley Demonstration Project (WVDP). This Program covers a number of activities that support high-level waste solidification, stabilization of facilities, and decontamination and decommissioning activities at the Project. The conduct of the Program provides confidence that all occupational radiation exposures received during operational tasks at the Project are within limits, standards, and program requirements, and are as low as reasonably achievable.
Date: May 6, 1997
Creator: Hoffman, R.L.
Partner: UNT Libraries Government Documents Department

Testing of the West Valley Vitrification Facility transfer cart control system

Description: Oak Ridge National Laboratory (ORNL) has designed and tested the control system for the West Valley Demonstration Project Vitrification Facility transfer cart. The transfer cart will transfer canisters of vitrified high-level waste remotely within the Vitrification Facility. The control system operates the cart under battery power by wireless control. The equipment includes cart-mounted control electronics, battery charger, control pendants, engineer`s console, and facility antennas. Testing was performed in several phases of development: (1) prototype equipment was built and tested during design, (2) board-level testing was then performed at ORNL during fabrication, and (3) system-level testing was then performed by ORNL at the fabrication subcontractor`s facility for the completed cart system. These tests verified (1) the performance of the cart relative to design requirements and (2) operation of various built-in cart features. The final phase of testing is planned to be conducted during installation at the West Valley Vitrification Facility.
Date: February 1995
Creator: Halliwell, J. W. & Bradley, E. C.
Partner: UNT Libraries Government Documents Department

Qualification testing and full-scale demonstration of titanium-treated zeolite for sludge wash processing

Description: Titanium-treated zeolite is a new ion-exchange material that is a variation of UOP (formerly Union Carbide) IONSIV IE-96 zeolite (IE-96) that has been treated with an aqueous titanium solution in a proprietary process. IE-96 zeolite, without the titanium treatment, has been used since 1988 in the West Valley Demonstration Project`s (WVDP) Supernatant Treatment System (STS) ion-exchange columns to remove Cs-137 from the liquid supernatant solution. The titanium-treated zeolite (TIE-96) was developed by Battelle-Pacific Northwest Laboratory (PNL). Following successful lab-scale testing of the PNL-prepared TIE-96, UOP was selected as a commercial supplier of the TIE-96 zeolite. Extensive laboratory tests conducted by both the WVDP and PNL indicate that the TIE-96 will successfully remove comparable quantities of Cs-137 from Tank 8D-2 high-level radioactive liquid as was done previously with IE-96. In addition to removing Cs-137, TIE-96 also removes trace quantities of Pu, as well as Sr-90, from the liquid being processed over a wide range of operating conditions: temperature, pH, and dilution. The exact mechanism responsible for the Pu removal is not fully understood. However, the Pu that is removed by the TIE-96 remains on the ion-exchange column under anticipated sludge wash processing conditions. From May 1988 to November 1990, the WVDP processed 560,000 gallons of liquid high-level radioactive supernatant waste stored in Tank 8D-2. Supernatant is an aqueous salt solution comprised primarily of soluble sodium salts. The second stage of the high-level waste treatment process began November 1991 with the initiation of sludge washing. Sludge washing involves the mixing of Tank 8D-2 contents, both sludge and liquid, to dissolve the sulfate salts present in the sludge. Two sludge washes were required to remove sulfates from the sludge.
Date: June 30, 1997
Creator: Dalton, W.J.
Partner: UNT Libraries Government Documents Department