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Analysis of debris vacuumed from K-Reactor tank

Description: During the ultrasonic (UT) inspection of K-Reactor tank by the Equipment Engineering Section in the spring of 1990, solid material (termed debris) was seen on the bottom of the tank. When the UT inspection was complete, a specially designed underwater vacuum cleaner was used to collect the accumulation at 17 monitor pin positions. This material was sent to SRL for characterization as an action item of the Reactor Corrosion Mitigation Committee. Acquisition of this debris provided an opportunity to obtain first-hand information about conditions within the tank that affect corrosion conditions and/or moderator chemistry. The purpose of this memorandum is to describe the results of the analyses and the implications of what was found.
Date: January 20, 1992
Creator: Baumann, E. W.
Partner: UNT Libraries Government Documents Department

Analysis of deposit from K-Reactor heat exchanger 4A

Description: Characterization of deposits from the reactor system provides a means of directly assessing corrosion and chemistry conditions within the system. The recent analysis of debris vacuumed from the bottom of K-Reactor tank provided information and reassurance about the conditions within the tank that would affect corrosion or moderator chemistry. Further opportunity for surveillance within the reactor system was recognized when solid deposits were found on the moderator side of the K-Reactor heat exchanger 4A that failed in December 1991. A sample of deposited material from the face of the tube sheet at the inlet end was removed under the direction of Equipment Engineering Section personnel. The material was analyzed by the Analytical Development Section by techniques used earlier for the K-tank debris. Elemental content was determined by Scanning Electron Microscopy/Energy Dispersive Spectroscopy (SEM/EDS). Total chlorine content was determined by neutron activation analysis. Crystalline components were identified by X-Ray diffraction, and radionuclidic content characterized by alpha pulse height analysis, beta counting, scintillation counting, and gamma spectroscopy. The purpose of this memorandum is to report the results of these analyses.
Date: March 20, 1992
Creator: Baumann, E.W.
Partner: UNT Libraries Government Documents Department

Analysis of deposit from K-Reactor heat exchanger 4A. Revision 1

Description: Characterization of deposits from the reactor system provides a means of directly assessing corrosion and chemistry conditions within the system. The recent analysis of debris vacuumed from the bottom of K-Reactor tank provided information and reassurance about the conditions within the tank that would affect corrosion or moderator chemistry. Further opportunity for surveillance within the reactor system was recognized when solid deposits were found on the moderator side of the K-Reactor heat exchanger 4A that failed in December 1991. A sample of deposited material from the face of the tube sheet at the inlet end was removed under the direction of Equipment Engineering Section personnel. The material was analyzed by the Analytical Development Section by techniques used earlier for the K-tank debris. Elemental content was determined by Scanning Electron Microscopy/Energy Dispersive Spectroscopy (SEM/EDS). Total chlorine content was determined by neutron activation analysis. Crystalline components were identified by X-Ray diffraction, and radionuclidic content characterized by alpha pulse height analysis, beta counting, scintillation counting, and gamma spectroscopy. The purpose of this memorandum is to report the results of these analyses.
Date: March 20, 1992
Creator: Baumann, E. W.
Partner: UNT Libraries Government Documents Department

Carbon-14 removal for disposal of reactor deionizer resins

Description: Disposal of depleted ion exchange resins from the primary system of the Savannah River Site (SRS) reactors is complicated by the presence of Carbon-14. Because Carbon-14 has a long half-life (5,730 years) and high mobility in soils, burial of the resins is no longer a viable option. Consequently some 35 spent reactor deionizers have accumulated that are to be stored aboveground in H-Area for an indefinite period. Spent deionizers containing Carbon-14 will continue to accumulate with operation of the present production reactors and would also accumulate from the proposed heavy water new production reactor. Removal of the Carbon-14 from the resins would reduce the volume of Carbon-14 bearing waste and enable the resins to be disposed of as low-level waste. Studies at SRS have indicated that the Carbon-14 from reactor primary coolant is mostly retained by the resins as the bicarbonate anion. Thus Carbon-14 removal might be accomplished by an acidification operation with trapping of the carbon dioxide released, for separate disposal. Conversion of the bicarbonate from the resin to barium carbonate, for example, would reduce the volume of waste more than a hundredfold. Displacement and recovery of Carbon-14 dioxide from reactor coolant deionizers by acid treatment has been reported by the Canadians. This memorandum recommends that a process be developed for Carbon-14 dioxide removal from SRS spent reactor deionizer resins, drawing on the Canadian experience.
Date: October 31, 1991
Creator: Carlton, W.H. & Baumann, E.W.
Partner: UNT Libraries Government Documents Department

Carbon-14 removal for disposal of reactor deionizer resins

Description: Disposal of depleted ion exchange resins from the primary system of the Savannah River Site (SRS) reactors is complicated by the presence of Carbon-14. Because Carbon-14 has a long half-life (5,730 years) and high mobility in soils, burial of the resins is no longer a viable option. Consequently some 35 spent reactor deionizers have accumulated that are to be stored aboveground in H-Area for an indefinite period. Spent deionizers containing Carbon-14 will continue to accumulate with operation of the present production reactors and would also accumulate from the proposed heavy water new production reactor. Removal of the Carbon-14 from the resins would reduce the volume of Carbon-14 bearing waste and enable the resins to be disposed of as low-level waste. Studies at SRS have indicated that the Carbon-14 from reactor primary coolant is mostly retained by the resins as the bicarbonate anion. Thus Carbon-14 removal might be accomplished by an acidification operation with trapping of the carbon dioxide released, for separate disposal. Conversion of the bicarbonate from the resin to barium carbonate, for example, would reduce the volume of waste more than a hundredfold. Displacement and recovery of Carbon-14 dioxide from reactor coolant deionizers by acid treatment has been reported by the Canadians. This memorandum recommends that a process be developed for Carbon-14 dioxide removal from SRS spent reactor deionizer resins, drawing on the Canadian experience.
Date: October 31, 1991
Creator: Carlton, W. H. & Baumann, E. W.
Partner: UNT Libraries Government Documents Department

Chemical aspects of gadolinium nitrate as soluble nuclear poison in Savannah River Plant reactors

Description: The aqueous solution chemistry of gadolinium nitrate was studied to identify conditions that interfere with successful cleanup of gadolinium in Savannah River Plant reactor systems. Injecting a gadolinium nitrate solution into the D/sub 2/O coolant-moderator constitutes a supplementary mode of reactor shutdown. The resulting approximately 0.001M gadolinium nitrate solution is then deionized by recirculation through mixed-bed ion exchange resins before reactor operation is resumed.
Date: January 1, 1978
Creator: Baumann, E.W.
Partner: UNT Libraries Government Documents Department

Colorimetric determination of Fe sup 2+ /Fe sup 3+ ratio in radioactive glasses

Description: In the vitrification of nuclear wastes, the Fe{sup 2+}/Fe{sup 3+} ratio in the glass is a measure of the redox properties of the glass melt. It is necessary to measure this ratio to ensure that the melt redox properties are suitable for the glass melter. A colorimetric method for measuring the Fe{sup 2+}/Fe{sup 3+} ratio in highly radioactive glasses was developed and tested remotely in a shielded cell. The tests were performed on glasses similar in composition and radioactivity to those that will be produced in the Savannah River Site Defense Waste Processing Facility. The first step of the method is dissolution of finely crushed glass with a hydrofluoric/sulfuric acid mixture with ammonium vanadate added to preserve the Fe{sup 2+} content of the glass during the dissolution. Boric acid is then added to complex fluoride and to destroy iron-fluoride complexes. After adjusting the solution to pH 5, FerroZine{sup TM} (trademark of the Hach Company, Loveland, CO) reagent is added to form a magenta-colored complex with Fe{sup 2+}. The absorbance at 562 nm is measured by using a fiber optic-coupled photodiode array spectrophotometer. Ascorbic acid is then used to reduce all the iron in solution to Fe{sup 2+} and the absorbance is again measured. The difference in absorbance measurements corresponds to the Fe{sup 3+} in the sample and the Fe{sup 2+}/Fe{sup 3+} ratio can be calculated.
Date: January 1, 1992
Creator: Coleman, C.J.; Baumann, E.W. & Bibler, N.E.
Partner: UNT Libraries Government Documents Department

Colorimetric determination of Fe{sup 2+}/Fe{sup 3+} ratio in radioactive glasses

Description: In the vitrification of nuclear wastes, the Fe{sup 2+}/Fe{sup 3+} ratio in the glass is a measure of the redox properties of the glass melt. It is necessary to measure this ratio to ensure that the melt redox properties are suitable for the glass melter. A colorimetric method for measuring the Fe{sup 2+}/Fe{sup 3+} ratio in highly radioactive glasses was developed and tested remotely in a shielded cell. The tests were performed on glasses similar in composition and radioactivity to those that will be produced in the Savannah River Site Defense Waste Processing Facility. The first step of the method is dissolution of finely crushed glass with a hydrofluoric/sulfuric acid mixture with ammonium vanadate added to preserve the Fe{sup 2+} content of the glass during the dissolution. Boric acid is then added to complex fluoride and to destroy iron-fluoride complexes. After adjusting the solution to pH 5, FerroZine{sup TM} (trademark of the Hach Company, Loveland, CO) reagent is added to form a magenta-colored complex with Fe{sup 2+}. The absorbance at 562 nm is measured by using a fiber optic-coupled photodiode array spectrophotometer. Ascorbic acid is then used to reduce all the iron in solution to Fe{sup 2+} and the absorbance is again measured. The difference in absorbance measurements corresponds to the Fe{sup 3+} in the sample and the Fe{sup 2+}/Fe{sup 3+} ratio can be calculated.
Date: May 1, 1992
Creator: Coleman, C. J.; Baumann, E. W. & Bibler, N. E.
Partner: UNT Libraries Government Documents Department

Colorimetric determination of low pH with malachite green. Revision 1

Description: A spectrophotometric method was developed for determination of concentration-based pH values from 0 to 2 with malachite green indicator. A quadratic model equation was based on the ratio of the absorbances of the peak at 618 nm and the isosbestic point at 518 nm. Normalization to the isosbestic point was used to stabilize the response because the color faded; the useful time interval was within 5 minutes after indicator addition. Model and verification sets agreed within {plus_minus}0.02 pH units between pH 0.3 and 1.8. This excellent precision makes the colorimetric method useful for acid determinations with a relative precision of >{plus_minus}5%. The presence of salts at a salt/acid equivalent ratio >0.l caused a low pH bias.
Date: April 1, 1994
Creator: Cloessner, P. F. & Baumann, E. W.
Partner: UNT Libraries Government Documents Department

Determination of Fe(II)Fe(II) ratio in glass

Description: The procedure was designed for the simple, rapid determination of the Fe(II)/Fe(III) ratio in glass samples. The procedure consists of the following steps: dissolution of the pulverized glass sample in a sulfuric-hydrofluoric acid mixture, containing ammonium vanadate, which preserves the Fe(II) content; addition of boric acid to destroy iron-fluoride complexes, making the iron available for color formation with Ferrozine; addition of pH 5 buffer and Ferrozine reagent to form the magenta-colored ferrous-Ferrozine complex, with measurement of the absorbance for the determination of Fe(II) content; and, addition of ascorbic acid to reduce Fe(III) to Fe(II), with a second absorbance measurement that determines total Fe. Directions for the preparation of glass from non-radioactive sludge samples are provided. The analysis of this prepared glass for the Fe(II)/Fe(III) ratio is an indication of the ratio that would be in a plant batch of glass if made from this sludge.
Date: July 26, 1989
Creator: Baumann, E.W.
Partner: UNT Libraries Government Documents Department

Determination of Free Acid by Standard Addition with Potassium Thiocyanate as Complexant

Description: A method is described for determination of free acid in solutions containing the hydrolyzable ions Al (III), Cr(III), Fe(III), Hg(II), Ni(II), Th(IV), and U(VI). The concentration of the sample is calculated either by solving three simultaneous Nernst equations, by the Gran plot procedure, or by means of a microprocessor pH meter. Molar concentrations of metal ion up to 2.5 times that of the acid can be tolerated. The method has been applied to analysis of nuclear processing solutions that contain Pu(III), in addition to the ions listed above.
Date: May 29, 2001
Creator: Baumann, E.W.
Partner: UNT Libraries Government Documents Department

Determination of Parts-Per-Million Cesium in Simulated Nuclear Waste with the Cesium-Selective Electrode

Description: Because the molybdophosphate electrode was not sufficiently sensitive, and the reliability of the ''liquid state'' electrode for routine analysis was uncertain, a cesium-selective electrode of the proven liquid membrane type was developed. This paper describes preparation and testing of a liquid membrane electrode that contains cesium tetraphenylboron dissolved in 4-ethylnitrobenzene.
Date: October 23, 2001
Creator: Baumann, E. W.
Partner: UNT Libraries Government Documents Department

Determination of Stability Constants of Hydrogen and Aluminum Fluorides with a Fluoride-Selective Electrode

Description: The ability to directly determine free fluoride ion concentration (or mean activity) simplifies gathering and interpretation of experimental data for studies of metal complexes. In this work, the new lanthanum fluoride electrode was used to measure free fluoride ion in an investigation of the hydrogen-fluoride and aluminum-fluoride systems in NH4NO3.
Date: January 6, 2003
Creator: Baumann, E.W.
Partner: UNT Libraries Government Documents Department

Effects of Gamma Radiation on Individual and Mixed Ion Exchange Resins

Description: The ion exchange resins that are used to deionize moderator in the reactor purification systems may accumulate sufficient radiation dose to damage the resins. This radiation damage would be manifested by: (1) loss of useful exchange capacity of the bed, which is costly since resins from the reactor deionizers are not reused; (2) shrinking or swelling of the resins, which may have some effect on the hydraulic behavior of the beds; (3) release of resin degradation products into the process stream, which pollutes moderator with impurities and precursors of the neutron-induced radioisotopes. This document details results of a laboratory study to determine the magnitude of these three effects by gamma irradiation of individual resins and their mixtures.
Date: January 6, 2003
Creator: Baumann, E.W.
Partner: UNT Libraries Government Documents Department

Evaluation of macroreticular anion exchange resin - RTA-893-R

Description: The macroreticular anion exchange resin, Amberlite IRA-900-OH, and an experimental resin from Ionac Chemical Company were irradiated by a /sup 60/Co source to doses of 5 x 10/sup 7/ rad and 10/sup 8/ rad. These doses approximate or exceed the dose encountered by the deionizer resins in 100-Area service. The loss in exchange capacity and the volume changes of Amberlite IRA-900 on irradiation were similar to those found previously for Amberlite IRA-400. The Ionac experimental resin, which had a considerably lower initial exchange capacity, likewise was not stable to radiation. It is concluded that Amberlite IRA-900 offers no advantage over Amberlite IRA-400 for 100-Area purification service. Since there is little justification for further evaluation of macroreticular resin for 100-Area use, the present work completes RTA-893-R.
Date: June 7, 1965
Creator: Baumann, E.W.
Partner: UNT Libraries Government Documents Department

Method for keeping RBOF waste within EPA pH limits for nonhazardous waste

Description: A system to control corrosion, while reducing the amount of hazardous waste generated, is proposed for the RBOF receiving basin. Details of the chemistry, experimental setup, and methods for monitoring pH are included in this report.
Date: February 18, 1992
Creator: Cobb, C.L.; Baumann, E.W.; Wilson, W.N. & Young, J.E.
Partner: UNT Libraries Government Documents Department

Method for keeping RBOF waste within EPA pH limits for nonhazardous waste

Description: A system to control corrosion, while reducing the amount of hazardous waste generated, is proposed for the RBOF receiving basin. Details of the chemistry, experimental setup, and methods for monitoring pH are included in this report.
Date: February 18, 1992
Creator: Cobb, C. L.; Baumann, E. W.; Wilson, W. N. & Young, J. E.
Partner: UNT Libraries Government Documents Department

Preparation and Properties of Nitrate-Deficient Gadolinium Nitrate Solutions

Description: Because of the high neutron absorption cross sections of some gadolinium isotopes, gadolinium salts in solution are used to control nuclear reactivity in aqueous systems. The present studies concern the preparation and analysis of nitrate-deficient solutions, the effect of time and gamma radiation on their stability, and the determination of the solubility of gadolinium hydroxide in H2O and D2O.
Date: March 15, 2001
Creator: Baumann, E.W.
Partner: UNT Libraries Government Documents Department

Reactor service life extension program

Description: A review of the Savannah River Site production reactor systems was initiated in 1980 and led to implementation of the Reactor Materials Program in 1984 to assess reactor safety and reactor service life. The program evaluated performance of the reactor tanks, primary coolant piping, and thermal shields, components of welded construction that were fabricated from Type 304 stainless steel. The structural integrity analysis of the primary coolant system has shown that the pressure boundary is not susceptible to gross rupture, including a double ended guillotine break or equivalent large area bank. Residual service life is potentially limited by two material degradation modes, irradiation damage and intergranular stress corrosion cracking. Analysis of the structural integrity of the tanks and piping has shown that continued safe operation of the reactors for several additional decades is not limited by the material performance of the primary coolant system. Although irradiation damage has not degraded material behavior to an unacceptable level, past experience has revealed serious difficulties with repair welding on irradiated stainless steel. Stress corrosion can be mitigated by newly identified limits on impurity concentrations in the coolant water and by stress mitigation of weld residual stresses. Work continues in several areas: the effects of helium on mechanical behavior of irradiated stainless steel; improved weld methods for piping and the reactor tanks; and a surveillance program to track irradiation effects on the tank walls.
Date: January 1, 1990
Creator: Caskey, G.R.; Sindelar, R.L.; Ondrejcin, R.S. & Baumann, E.W.
Partner: UNT Libraries Government Documents Department