46 Matching Results

Search Results

Advanced search parameters have been applied.

Report on Analyses of WAC Samples of Evaporator Overheads - 2004

Description: In November and December of 2004, the Tank Farm submitted annual samples from 2F, 2H and 3H Evaporator Overhead streams for characterization to verify compliance with the new Effluent Treatment Facility (ETF) Waste Acceptance Criteria (WAC) and to look for organic species. With the exception of slightly high ammonia in the 2F evaporator overheads and high radiation control guide number for the 3H and 2F evaporator overhead samples, all the overheads samples were found to be in compliance with the Effluent Treatment Facility WAC. The ammonium concentration in the 2F-evaporator overhead, at 33 mg/L, was above the ETF waste water collection tank (WWCT) limits of 28 mg/L. The RCG Number for the 3H and 2F evaporator samples at, respectively, 1.38E-02 and 8.24E-03 were higher than the WWCT limit of 7.69E-03. The analytical detection limits for americium-241 and radium-226 in the evaporator samples were not consistently met because of low WWCT detection limits and insufficient evaporator samples.
Date: March 18, 2005
Creator: Oji, L
Partner: UNT Libraries Government Documents Department

Analysis Of 2H-Evaporator Scale Pot Bottom Sample [HTF-13-11-28H]

Description: Savannah River Remediation (SRR) is planning to remove a buildup of sodium aluminosilicate scale from the 2H-evaporator pot by loading and soaking the pot with heated 1.5 M nitric acid solution. Sampling and analysis of the scale material from the 2H evaporator has been performed so that the evaporator can be chemically cleaned beginning July of 2013. Historically, since the operation of the Defense Waste Processing Facility (DWPF), silicon in the DWPF recycle stream combines with aluminum in the typical tank farm supernate to form sodium aluminosilicate scale mineral deposits in the 2H-evaporator pot and gravity drain line. The 2H-evaporator scale samples analyzed by Savannah River National Laboratory (SRNL) came from the bottom cone sections of the 2H-evaporator pot. The sample holder from the 2H-evaporator wall was virtually empty and was not included in the analysis. It is worth noting that after the delivery of these 2H-evaporator scale samples to SRNL for the analyses, the plant customer determined that the 2H evaporator could be operated for additional period prior to requiring cleaning. Therefore, there was no need for expedited sample analysis as was presented in the Technical Task Request. However, a second set of 2H evaporator scale samples were expected in May of 2013, which would need expedited sample analysis. X-ray diffraction analysis (XRD) confirmed the bottom cone section sample from the 2H-evaporator pot consisted of nitrated cancrinite, (a crystalline sodium aluminosilicate solid), clarkeite and uranium oxide. There were also mercury compound XRD peaks which could not be matched and further X-ray fluorescence (XRF) analysis of the sample confirmed the existence of elemental mercury or mercuric oxide. On ''as received'' basis, the scale contained an average of 7.09E+00 wt % total uranium (n = 3; st.dev. = 8.31E-01 wt %) with a U-235 enrichment of 5.80E-01 % (n = 3; ...
Date: July 15, 2013
Creator: Oji, L. N.
Partner: UNT Libraries Government Documents Department

Analysis Of 2H-Evaporator Scale Wall [HTF-13-82] And Pot Bottom [HTF-13-77] Samples

Description: Savannah River Remediation (SRR) is planning to remove a buildup of sodium aluminosilicate scale from the 2H-evaporator pot by loading and soaking the pot with heated 1.5 M nitric acid solution. Sampling and analysis of the scale material has been performed so that uranium and plutonium isotopic analysis can be input into a Nuclear Criticality Safety Assessment (NCSA) for scale removal by chemical cleaning. Historically, since the operation of the Defense Waste Processing Facility (DWPF), silicon in the DWPF recycle stream combines with aluminum in the typical tank farm supernate to form sodium aluminosilicate scale mineral deposits in the 2H-evaporator pot and gravity drain line. The 2H-evaporator scale samples analyzed by Savannah River National Laboratory (SRNL) came from two different locations within the evaporator pot; the bottom cone sections of the 2H-evaporator pot [Sample HTF-13-77] and the wall 2H-evaporator [sample HTF-13-82]. X-ray diffraction analysis (XRD) confirmed that both the 2H-evaporator pot scale and the wall samples consist of nitrated cancrinite (a crystalline sodium aluminosilicate solid) and clarkeite (a uranium oxyhydroxide mineral). On ''as received'' basis, the bottom pot section scale sample contained an average of 2.59E+00 {+-} 1.40E-01 wt % total uranium with a U-235 enrichment of 6.12E-01 {+-} 1.48E-02 %, while the wall sample contained an average of 4.03E+00 {+-} 9.79E-01 wt % total uranium with a U-235 enrichment of 6.03E-01% {+-} 1.66E-02 wt %. The bottom pot section scale sample analyses results for Pu-238, Pu-239, and Pu-241 are 3.16E-05 {+-} 5.40E-06 wt %, 3.28E-04 {+-} 1.45E-05 wt %, and <8.80E-07 wt %, respectively. The evaporator wall scale samples analysis values for Pu-238, Pu-239, and Pu-241 averages 3.74E-05 {+-} 6.01E-06 wt %, 4.38E-04 {+-} 5.08E-05 wt %, and <1.38E-06 wt %, respectively. The Pu-241 analyses results, as presented, are upper limit values. For these two evaporator scale samples obtained ...
Date: September 11, 2013
Creator: Oji, L. N.
Partner: UNT Libraries Government Documents Department

Tank 30 and 37 Supernatant Sample Cross-Check and Evaporator Feed Qualification Analysis-2012

Description: This report summarizes the analytical data reported by the F/H and Savannah River National Laboratories for the 2012 cross-check analysis for high level waste supernatant liquid samples from SRS Tanks 30 and 37. The intent of this Tank 30 and 37 sample analyses was to perform cross-checks against routine F/H Laboratory analyses (corrosion and evaporator feed qualification programs) using samples collected at the same time from both tanks as well as split samples from the tanks.
Date: March 7, 2013
Creator: Oji, L. N.
Partner: UNT Libraries Government Documents Department

ANALYSIS OF 2H-EVAPORATOR SCALE WALL [HTF-13-82] AND POT BOTTOM [HTF-13-77] SAMPLES

Description: Savannah River Remediation (SRR) is planning to remove a buildup of sodium aluminosilicate scale from the 2H-evaporator pot by loading and soaking the pot with heated 1.5 M nitric acid solution. Sampling and analysis of the scale material has been performed so that uranium and plutonium isotopic analysis can be input into a Nuclear Criticality Safety Assessment (NCSA) for scale removal by chemical cleaning. Historically, since the operation of the Defense Waste Processing Facility (DWPF), silicon in the DWPF recycle stream combines with aluminum in the typical tank farm supernate to form sodium aluminosilicate scale mineral deposits in the 2Hevaporator pot and gravity drain line. The 2H-evaporator scale samples analyzed by Savannah River National Laboratory (SRNL) came from the bottom cone sections of the 2H-evaporator pot [Sample HTF-13-77] and the wall 2H-evaporator [sample HTF-13-82]. X-ray diffraction analysis (XRD) confirmed that both the 2H-evaporator pot scale and the wall samples consist of nitrated cancrinite (a crystalline sodium aluminosilicate solid) and clarkeite (a uranium oxy-hydroxide mineral). On “as received” basis, the bottom pot section scale sample contained an average of 2.59E+00 ± 1.40E-01 wt % total uranium with a U-235 enrichment of 6.12E-01 ± 1.48E-02 %, while the wall sample contained an average of 4.03E+00 ± 9.79E-01 wt % total uranium with a U-235 enrichment of 6.03E-01% ± 1.66E-02 wt %. The bottom pot section scale sample analyses results for Pu-238, Pu-239, and Pu-241 are 3.16E- 05 ± 5.40E-06 wt %, 3.28E-04 ± 1.45E-05 wt %, and &lt;8.80E-07 wt %, respectively. The evaporator wall scale samples analysis values for Pu-238, Pu-239, and Pu-241 averages 3.74E-05 ± 6.01E-06 wt %, 4.38E-04 ± 5.08E-05 wt %, and &lt;1.38E-06 wt %, respectively. The Pu-241 analyses results, as presented, are upper limit values. These results are provided so that SRR can calculate the equivalent uranium-235 concentrations ...
Date: June 21, 2013
Creator: Oji, L.
Partner: UNT Libraries Government Documents Department

Characterization of Tank 23H Supernate Per Saltstone Waste Acceptance Criteria Analysis Requirements-2005

Description: Variable depth Tank 23H samples (22-inch sample [HTF-014] and 185-inch sample [HTF-013]) were pulled from Tank 23H in February, 2005 for characterization. The characterization of the Tank 23H low activity waste is part of the overall liquid waste processing activities. This characterization examined the species identified in the Saltstone Waste Acceptance Criteria (WAC) for the transfer of waste into the Salt-Feed Tank (SFT). The samples were delivered to the Savannah River National Laboratory (SRNL) and analyzed. Apart from radium-226 with an average measured detection limit of &lt; 2.64E+03 pCi/mL, which is about the same order of magnitude as the WAC limit (&lt; 8.73E+03 pCi/mL), none of the species analyzed was found to approach the limits provided in the Saltstone WAC. The concentration of most of the species analyzed for the Tank 23H samples were 2-5 orders of magnitude lower than the WAC limits. The achievable detection limits for a number of the analytes were several orders of magnitude lower than the WAC limits, but one or two orders of magnitude higher than the requested detection limits. Analytes which fell into this category included plutonium-241, europium-154/155, antimony-125, tin-126, ruthenium/rhodium-106, selenium-79, nickel-59/63, ammonium ion, copper, total nickel, manganese and total organic carbon.
Date: June 1, 2005
Creator: Oji, L
Partner: UNT Libraries Government Documents Department

Characterization of Tank 23H Supernate Per Saltstone Waste Acceptance Criteria Analysis Requirements -2005

Description: Variable depth Tank 23H samples (22-inch sample [HTF-014] and 185-inch sample [HTF-013]) were pulled from Tank 23H in February, 2005 for characterization. The characterization of the Tank 23H low activity waste is part of the overall liquid waste processing activities. This characterization examined the species identified in the Saltstone Waste Acceptance Criteria (WAC) for the transfer of waste into the Salt-Feed Tank (SFT). The samples were delivered to the Savannah River National Laboratory (SRNL) and analyzed. Apart from radium-226 with an average measured detection limit of &lt; 2.64E+03 pCi/mL, which is about the same order of magnitude as the WAC limit (&lt; 8.73E+03 pCi/mL), none of the species analyzed was found to approach the limits provided in the Saltstone WAC. The concentration of most of the species analyzed for the Tank 23H samples were 2-5 orders of magnitude lower than the WAC limits. The achievable detection limits for a number of the analytes were several orders of magnitude lower than the WAC limits, but one or two orders of magnitude higher than the requested detection limits. Analytes which fell into this category included plutonium-241, europium-154/155, antimony-125, tin-126, ruthenium/rhodium-106, selenium-79, nickel-59/63, ammonium ion, copper, total nickel, manganese and total organic carbon.
Date: May 5, 2005
Creator: Oji, L
Partner: UNT Libraries Government Documents Department

Oxidative Mineralization and Characterization of Polyvinyl Alcohol Solutions for Wastewater Treatment

Description: Photochemical and ultrasonic treatment of polyvinyl alcohol (PVA), derived from PVA fabric material, with hydrogen peroxide was evaluated as a primary method for PVA mineralization into simpler organic molecules. PVA-based waste streams have been found to be compatible with nuclear process wastewater treatment facilities only when solubilized PVA is more than 90 percent mineralized with hydrogen peroxide. No undesirable solid particles are formed with other nuclear process liquid waste when they are mixed, pH adjusted, evaporated and blended with this type of oxidized PVA waste streams. The presence of oxidized PVA in a typical nuclear process wastewater has been found to have no detrimental effect on the efficiency of ion exchange resins, inorganic, and precipitation agents used for the removal of radionuclides from nuclear waste streams. The disappearance of PVA solution in hydrogen peroxide with ultrasonic/ ultraviolet irradiation treatment was characterized by pseudo-first-order reaction kinetics. Radioactive waste contaminated PVA fabric can be solubilized and mineralized to produce processible liquid waste, hence, no bulky solid waste disposal cost can be incurred and the radionuclides can be effectively recovered. Therefore, PVA fabric materials can be considered as an effective substitute for cellulose fabrics that are currently used in radioactive waste decontamination processes.
Date: August 7, 2003
Creator: Oji, L.N.
Partner: UNT Libraries Government Documents Department

The Evaluation of Uranium-236 Isotopic Dilution with the Addition of Depleted Uranium to Supernatant Liquid Waste

Description: This paper describes laboratory-scale results on experiments performed to examine the feasibility of isotopic dilution of uranium-235 in supernatant liquid storage tanks at the Savannah River Site. The isotopic dilution tests were accomplished by adding an alkaline depleted uranium solution to small portions of simulated and actual storage tank waste solutions with enriched U-235 compositions. Based on the laboratory observations, recommendations were made, which involved the addition of significant quantities of uranyl carbonate solution to over four million liters of U-235 enriched waste stored in Tank 43H at SRS to reduce the risk for criticality. A post-uranyl carbonate addition analysis on the tank supernate confirmed the effectiveness of depleted uranium in isotopic dilution of U-235. The U-235 enrichment in the Tank 43H was isotopically diluted from an original high of over 4 wt percent down to less than 0.5 wt percent as predicted from the laboratory investigations.
Date: June 27, 2003
Creator: Oji, L.N.
Partner: UNT Libraries Government Documents Department

Analysis of Tank 43H Samples at the Conclusion of Uranyl Carbonate Addition

Description: Tank 43H serves as the feed Tank to the 2H evaporator. In the months of July and August 2001, about 21,000 gallons of a depleted uranyl carbonate solution were added to Tank 43H and agitated with two Flygt mixers. The depleted uranium addition served to decrease the U-235 enrichment in the Tank 43H supernate so that the supernate could be evaporated with no risk of accumulating enriched uranium.
Date: October 15, 2002
Creator: Oji, L.N.
Partner: UNT Libraries Government Documents Department

Zinc Bromide Combustion: Implications for the Consolidated Incinerator Facility

Description: In the nuclear industry, zinc bromide (ZnBr2) is used for radiation shielding. At Savannah River Site (SRS) zinc bromide solution, in appropriate configurations and housings, was used mainly for shielding in viewing windows in nuclear reactor and separation areas. Waste stream feeds that will be incinerated at the CIF will occasionally include zinc bromide solution/gel matrices.The CIF air pollution systems control uses a water-quench and steam atomizer scrubber that collects salts, ash and trace metals in the liquid phase. Water is re-circulated in the quench unit until a predetermined amount of suspended solids or dissolved salts are present. After reaching the threshold limit, "dirty liquid", also called "blowdown", is pumped to a storage tank in preparation for treatment and disposal. The air pollution control system is coupled to a HEPA pre-filter/filter unit, which removes particulate matter from the flue gas stream (1).The objective of this report is to review existing literature data on the stability of zinc bromide (ZnBr2) at CIF operating temperatures (>870 degrees C (1600 degrees F) and determine what the combustion products are in the presence of excess air. The partitioning of the combustion products among the quencher/scrubber solution, bottom ash and stack will also be evaluated. In this report, side reactions between zinc bromide and its combustion products with fuel oil were not taken into consideration.
Date: December 16, 1998
Creator: Oji, L.N.
Partner: UNT Libraries Government Documents Department

Ion Exchange Conceptual Design for Treating Seven Technical Area Sumps with Elevated Levels of Copper and Zinc

Description: Recently a meeting was held to discuss technical support for developing a conceptual design and estimate for installing and operating an in-line ion exchange system to treat seven Technical Area Sumps with elevated levels (high ppb - low ppm) of copper and potentially zinc (copper level is above the outfall limits). These sump waters are currently routed to the A01 outfall, which is permitted by the State of South Carolina. a study of potential treatment options and followup laboratory work done in the summer of 1997 by Larry Oji and John Hage identified two commercially available ion exchange resins, Duolit GT-73 and Chelex 100, for treating waters at these metals concentrations.
Date: February 17, 1999
Creator: Oji, L.N.
Partner: UNT Libraries Government Documents Department

Copper Removal from A-01 Outfall by Ion Exchange

Description: Chelex100, a commercially available ion exchange resin, has been identified in this study as having a significant affinity for copper and zinc in the A-01 outfall water. Removal of copper and zinc from A-01 outfall water will ensure that the outfall meets the state of South Carolina's limit on these heavy metals.
Date: February 17, 1999
Creator: Oji, L.N.
Partner: UNT Libraries Government Documents Department

Evaluation of the Radiation Stability of SuperLig 639

Description: A method for treatment and disposal of the Hanford High Level Waste has been proposed for BNFL, Inc. In this process, a portion of the Hanford High Level Waste will be pretreated to concentrate radionuclides prior to vitrification. This task examines the stability of one of the ion exchange resins, SuperLig (TM) 639, toward irradiation. These tests were conducted using simulated Hanford High Level Waste containing pertechnetate ion as a stand-in for pertechnetate.
Date: July 26, 2001
Creator: Oji, L.N.
Partner: UNT Libraries Government Documents Department

Oxidative mineralization and characterization of polyvinyl alcohol for compatibility with tank farm processing chemistry

Description: Polyvinyl alcohol (PVA) material has been evaluated for use as a cost-effective substitute for conventional cellulose-based disposal materials (decontamination mops and wipes), plastic bags, and disposable personal protection clothing, that are currently used at Savannah River Site. This study also provides process design criteria for ultraviolet/ultrasonic/hydrogen peroxide PVA reactor system.
Date: January 4, 2000
Creator: Oji, L.N.
Partner: UNT Libraries Government Documents Department

Evaluation of Crystalline Silicotitanate and Self-Assembled Monolayers on Mesoporous Support for Cesium and Mercury Removal from DWPF Recycle

Description: The affinities for cesium and mercury ions contained in DWPF recycle simulants and Tank-22H waste have been evaluated using Crystalline Silicotitanate (CST) and Self-Assembled Monolayers on Mesoporous Support (SAMMS) ion-exchange materials, respectively. Results of the performance evaluations of CST on the uptake of cesium with simulants and actual DWPF recycle samples (Tank 22H) indicate that, in practice, this inorganic ion-exchange material can be used to remove radioactive cesium from the DWPF recycle. SAMMS material showed little or no affinity for mercury from highly alkaline DWPF waste. However, at near neutral conditions (DWPF simulant solution pH adjusted to 7), SAMMS was found to have a significant affinity for mercury. Conventional Duolite/256 ion exchange material showed an increase in affinity for mercury with increase in DWPF recycle simulant pH. Duolite/256 GT-73 also had a high batch distribution coefficient for mercury uptake from actual Tank 22H waste.
Date: November 5, 1999
Creator: Oji, L.N.
Partner: UNT Libraries Government Documents Department

Evaluation of Uranium Co-precipitations with Sodium Aluminosilicate Phases

Description: This paper describes batch laboratory experiments performed to evaluate uranium incorporation into aluminosilicate structures during synthesis. This research was conducted in response to plant problems related to the accumulation of uranium with aluminosilicates in low-level radioactive waste evaporators. We have found that conditions which favor precipitation of aluminosilicates also foster uranium solid precipitation, so it is difficult to attribute problems with uranium accumulation to say just the formation of the aluminosilicates. Infrared spectra shows that sodium uranates, uranium silicates and other uranium solids are formed during the synthesis of sodium aluminosilicates structures in the presence of uranium. Both amorphous and sodalite aluminosilcate phases, unlike zeolite A phase, show appreciable affinity for uranium incorporation during their formation in the presence of uranium.
Date: June 24, 2003
Creator: Oji, L.N.
Partner: UNT Libraries Government Documents Department

Cesium Removal from the Fuel Storage Water at the Savannah River Site R-Building Disassembly Basin Using 3M Empore(r)-Membrane Filter Technology

Description: This report describes results from a seven-day demonstration of the use of 3M Empore(r) membrane filter loaded with ion exchange material potassium cobalt hexacynoferrate (CoHex) for cesium uptake from the R-Disassembly Basin at the Savannah River Site.
Date: October 4, 1999
Creator: Oji, L.N.
Partner: UNT Libraries Government Documents Department

Oxidative Mineralization and Characterization of Polyvinyl Alcohol Solutions for Wastewater Treatment

Description: The principal objectives of this study are to identify an appropriate polyvinyl alcohol (PVA) oxidative mineralization technique, perform compatibility and evaporation fate tests for neat and mineralized PVA, and determine potential for PVA chemical interferences which may affect ion exchange utilization for radioactive wastewater processing in the nuclear industry.
Date: August 31, 1999
Creator: Oji, L.N.
Partner: UNT Libraries Government Documents Department

Evaluation of the Incorporation of Uranium into Sodium Aluminosilicate Phases

Description: This report describes batch laboratory experiments performed to determine the relative amounts of uranium incorporated in aluminosilicate structures during synthesis. The findings summarized here are based on laboratory experiments, which involved the synthesis of sodium aluminosilicates (NAS) structures, amorphous, zeolites A and sodalite phases in the presence of depleted uranium and the analytical search for incorporated uranium in NAS internal structures after synthesis. These studies will support the basis for continued operation of evaporators at the Savannah River Site (SRS).
Date: March 26, 2003
Creator: Oji, L.N.
Partner: UNT Libraries Government Documents Department

Characterization of Tank 46F Core Samples

Description: Tank 46F is the current drop tank for the 2F evaporator and contains about190 inches of saltcake. Samples from the High Level Waste (HLW) Evaporator feed and drop tanks are analyzed every six months for criticality safety and scale formation potential. Analysis performed under this program includes analysis for silicon, aluminum, sodium, and free hydroxide concentration (evaluation of scale formation rates) and suite of criticality analyses . The recent minor leaks detected in Tank 5F have also led to HLW engineering to attempt to decide where to send material from Tank 5F. One of the primary options is to transfer Tank 5F material to Tank 46F. In order to support the Tank 5F transfer decision process, data is needed on the uranium isotopic distribution in the salt in Tank 46F.
Date: July 31, 2002
Creator: Oji, L.N.
Partner: UNT Libraries Government Documents Department

Evaluation of Absorbents for Compatibility with Site Generated Hazardous and Mixed Liquid Wastes

Description: SRS Solid Waste requested SRTC to perform a literature-based evaluation of sorbents, which are compatible with hazardous mixed waste being generated on site. Polypropylene-based materials and ground corn cob (Toxi-dry), because of their compatibility with the Consolidated Incinerator Facility (CIF) process, are the only two spill stabilization agents which are recommended for use on site (IS manual, Waste Acceptance Criteria 3.18). While ensuring minimal potential for undesired reactions between spills and spill control agents, Solid Waste wants to increase the number of site approved absorbents to give waste generators more flexibility in choosing liquid spill immobilization agents.
Date: March 12, 2002
Creator: Oji, L. N.
Partner: UNT Libraries Government Documents Department