Savannah River Technology Center Research Articles

Calculation of 237Np and 241Am detector calibration constants from first principles

The Analytical Development Section of Savannah River Technology Center (SRTC) was requested by the Facilities Disposition Projects (FDP) to determine the holdup of enriched uranium in the 321-M facility as part of an overall deactivation project of the facility. The 321-M facility was used to fabricate enriched uranium fuel assemblies, lithium-aluminum target tubes, neptunium assemblies, and miscellaneous components for the production reactors. The results of the holdup assays are essential for determining compliance with the Waste Acceptance Criteria, Material Control & Accountability, and to meet criticality safety controls. This report covers calibration of the detectors in order to support holdup measurements in the C and D out-gassing ovens. These ovens were used to remove gas entrained in billet assembly material prior to the billets being extruded into rods by the extrusion press. A portable high purity germanium detection system was used to determine highly enriched uranium (HEU) holdup and to determine holdup of 235U, 237Np, and 241Am that were observed in these components. The detector system was run by an EG&G Dart™ system that contains the high voltage power supply and signal processing electronics. A personal computer with Gamma-Vision software was used to control the Dart™ MCA and provide space to store and manipulate multiple 4096-channel γ-ray spectra. The measured 237Np and 241Am contents were especially important in these components because their presence is unusual and unexpected in 321-M. It was important to obtain a measured value of these two components to disposition the out-gassing ovens and to determine whether a separate waste stream was necessary for release of these contaminated components to the E-Area Solid Waste Vault. This report presents determination of the calibration constants from first principles for determination of 241Am and 237Np using this detection system and compares the values obtained for 237Np with the calibration factors obtained with a subsequent measurement using a point source of radioactive equilibrium 237Np/233Pa.

Laser scanning confocal microscopic investigations of simulated nuclear waste structures

Researchers at the Department of Energy's Savannah River Technology Center are using advanced microscopy techniques to understand the effects of trace organic chemical additions on nuclear waste slurry flow properties. Trace organic chemicals, surfactants (rheology modifiers), are being used in all types of industries to modify the flow properties of various commercial chemicals. Nuclear waste treatment at the Department of Energy's weapons production facilities, Savannah River Site and Hanford Reservation, is limited by the viscosity of the nuclear waste slurries as the material is processed through a variety of waste treatment and immobilization processes. The picture was taken using a laser scanning confocal microscope.

Calculation of 237Np and 241Am detector calibration constants from first principles

The Analytical Development Section of Savannah River Technology Center (SRTC) was requested by the Facilities Disposition Projects (FDP) to determine the holdup of enriched uranium in the 321-M facility as part of an overall deactivation project of the facility. The 321-M facility was used to fabricate enriched uranium fuel assemblies, lithium-aluminum target tubes, neptunium assemblies, and miscellaneous components for the production reactors. The results of the holdup assays are essential for determining compliance with the Waste Acceptance Criteria, Material Control & Accountability, and to meet criticality safety controls. This report covers calibration of the detectors in order to support holdup measurements in the C and D out-gassing ovens. These ovens were used to remove gas entrained in billet assembly material prior to the billets being extruded into rods by the extrusion press. A portable high purity germanium detection system was used to determine highly enriched uranium (HEU) holdup and to determine holdup of 235U, 237Np, and 241Am that were observed in these components. The detector system was run by an EG&G DartO system that contains the high voltage power supply and signal processing electronics. A personal computer with Gamma-Vision software was used to control the DartO MCA and provide space to store and manipulate multiple 4096-channel g-ray spectra. The measured 237Np and 241Am contents were especially important in these components because their presence is unusual and unexpected in 321-M. It was important to obtain a measured value of these two components to disposition the out-gassing ovens and to determine whether a separate waste stream was necessary for release of these contaminated components to the E-Area Solid Waste Vault. This report presents determination of the calibration constants from first principles for determination of 241Am and 237Np using this detection system and compares the values obtained for 237Np with the calibration factors obtained with a subsequent measurement using a point source of radioactive equilibrium 237Np/233Pa.

The Design and Fabrication of the Savannah River Technology Center's New Glovebox Contained Scanning Electron Microscope

Journal Article The Design and Fabrication of the Savannah River Technology Center's New Glovebox Contained Scanning Electron Microscope Get access Michael E Summer Michael E Summer Westinghouse Savannah River Site, Analytical Development Section, 773-A Room C-136, Aiken, SC 29808 Search for other works by this author on: Oxford Academic Google Scholar Microscopy and Microanalysis, Volume 9, Issue S02, 1 August 2003, Pages 1048–1049, https://doi.org/10.1017/S1431927603445248 Published: 24 July 2003

Technetium Removal from Hanford and Savannah River Site Actual Tank Waste Supernates with Superlig® 639 Resin

SuperLig 639 elutable, organic resin was selected for technetium (as pertechnetate ion) removal from Hanford Site radioactive waste samples as part of the River Protection Project—Waste Treatment Plant (RPP-WTP) design. In support of the RPP-WTP flow sheet development, column tests have been performed at the Savannah River Technology Center with SuperLig 639 resin using actual Hanford Site tank waste samples. The resin was shown to be highly effective at pertechnetate removal from these caustic, high-sodium, aqueous waste samples. Pertechnetate ion was subsequently eluted from the columns with water. An additional column test conducted on a Savannah River Site waste sample revealed exceptional performance, presumably due to the fact that lower concentrations of competing anions (primarily nitrate) were present in the sample.

Cross-flow filtration with a shear-thinning organic-based slurry

The Department of Energy is sponsoring the River Protection Project, which includes the design of a facility to stabilize liquid radioactive waste that is stored at the Hanford Site. The Savannah River Technology Center (SRTC) of the Westinghouse Savannah River Company was contracted to develop and test parts of the waste treatment process. One part of the process is the separation of highly radioactive solids from the liquid wastes by precipitation and cross-flow filtration. A cross-flow filter was tested with simulated wastes made to represent typical waste chemical and physical characteristics. This paper discusses the results of a cross-flow filter operation in a pilot-scale facility that was designed, built, and run by the Experimental Thermal Fluids Laboratory of SRTC. The waste simulant had an insoluble solids loading from 2 to 22 wt.%, with particle sizes from 1 to 2 μm and 5 to 10 μm in diameter. The filter contained seven 316L stainless steel sintered-metal tubes in parallel. Each tube was identical, being 1.01 m long, 9.5 mm inside diameter, and membrane thickness of 1.6 mm. The porous tubes were made to capture 95% of particles ⩾0.1 μm. The flow conditions for the test varied: Axial slurry velocities from 3 to 5 m/s and transmembrane pressures (TMP) from 200 to 500 kPa, at a temperature of 298 K. The results showed good separation for the Newtonian slurry with solids loadings of 2 wt.%. Under the flow conditions of a slurry velocity of 3.5 m/s and a TMP of 400 kPa, the observed filtrate flux was 17 cm/h (Volumetric flow rate per filter surface area (cm 3/h/cm 2) was reduced to cm/h for convenience. Other commonly used units for filter flux include: 1 cm/h=10 l/m 2/h=0.0041 gpm/ft 2.); meeting the plant design need of 16 cm/h. As the insoluble solids loading was increased to 22 wt.% the slurry became a shear thinning, thixotropic fluid, with a yield stress of 5.5 Pa and a consistency of 23 mPa s. Despite the changing rheology the filtrate flux continued, but dropped to 2.4 cm/h.

Effects of hoisting on the input shaping control of gantry cranes

The dynamic behavior of a planar gantry crane with hoisting of the load is investigated. The command generation method of input shaping is proposed for reduction of the residual vibration. Several versions of input shaping are evaluated and compared with time-optimal rigid-body commands over a wide range of parameters. Input shaping provides significant reduction in both the residual and transient oscillations, even when the hoisting distance is a large percentage of the cable length. Experimental results from a 15-ton gantry crane at the Savannah River Technology Center are used to support the numerical results.

Characterization of and Waste Acceptance Radionuclides to be Reported for the Second Macro-Batch of High-Level Waste Sludge Being Vitrified in the DWPF Melter

The Defense Waste Processing Facility (DWPF), at the Savannah River Site (SRS), is currently processing the second million gallon batch (Macro-Batch 2) of radioactive sludge slurry into a durable borosilicate glass for permanent geological disposal. To meet the reporting requirements as specified in the Department of Energy's Waste Acceptance Product Specifications (WAPS), for the final glass product, the nonradioactive and radioactive compositions must be provided for a Macro-Batch of material. In order to meet this requirement, sludge slurry samples from Macro-Batch 2 were analyzed in the Shielded Cells Facility of the Savannah River Technology Center (SRTC). This information is used to complete the necessary Production Records at DWPF so that the final glass product, resulting from Macro Batch 2, may be disposed of at a Federal Repository. This paper describes the results obtained from the analyses of the sludge slurry samples taken from Macro-Batch 2 to meet the reporting requirements of the WAPS. Twenty eight elements were identified for the nonradioactive composition and thirty one for the radioactive composition. The reportable radioisotopes range from C-14 to Cm-246.

Zero dead time spectroscopy without full charge collection

The Savannah River Technology Center has built a remote gamma monitoring instrument which employs data sampling techniques rather than full charge collection to perform energy spectroscopy without instrument dead time. The raw, unamplified anode output of a photomultiplier tube is directly coupled to the instrument to generate many digital samples during the charge collection process, so that all pulse processing is done in the digital domain. The primary components are a free running, 32 MSPS, 10-bit A/D, a field programmable gate array, FIFO buffers, and a digital signal processor (DSP). Algorithms for pulse integration, pile-up rejection, and other shape-based criteria are being developed in DSP code for migration into the gate array. Spectra taken with a 2 in NaI(Tl) detector have been obtained at rates as high as 59 000 counts per second without dead time with peak resolution at 662 keV measuring 7.3%.

Immobilization of rocky flats graphite fines residues

AbstractThe Savannah River Technology Center (SRTC) is developing an immobilization process for graphite fines residues generated during nuclear materials production activities at the Rocky Flats Environmental Technology Site (Rocky Flats). The continued storage of this material has been identified as an item of concern. The residue was generated during the cleaning of graphite casting molds and potentially contains reactive plutonium metal. The average residue composition is 73 wt% graphite, 15 wt% calcium fluoride (CaF2), and 12 wt% plutonium oxide (PuO2 ). Approximately 950 kg of this material are currently stored at Rocky Flats.The strategy of the immobilization process is to microencapsulate the residue by mixing with a sodium borosilicate (NBS) glass frit and heating at nominally 700°C. The resulting waste form would be sent to the Waste Isolation Pilot Plant (WIPP) for disposal. Since the PuO2 concentration in the residue averages 12 wt%, the immobilization process was required to meet the intent of safeguards termination criteria by limiting plutonium recoverability based on a test developed by Rocky Flats. The test required a plutonium recovery of less than 4 g/kg of waste form when a sample was leached using a nitric acid/CaF2 dissolution flowsheet.Immobilization experiments were performed using simulated graphite fines with cerium oxide (CeO2) as a surrogate for PuO2 and with actual graphite fines residues. Small-scale surrogate experiments demonstrated that a 4:1 frit to residue ratio was adequate to prevent recovery of greater than 4 g/kg of cerium from simulated waste forms. Additional experiments investigated the impact of varying concentrations of CaF2 and the temperature/heating time cycle on the cerium recovery. Optimal processing conditions developed during these experiments were subsequently demonstrated at full-scale with surrogate materials and on a smaller scale using actual graphite fines.In general, the recovery of cerium from the full-scale waste forms was higher than for smaller scale experiments. The presence of CaF2 also caused a dramatic increase in cerium recovery not seen in the small-scale experiments. However, the results from experiments with actual graphite fines were encouraging. A 4:1 frit to residue ratio, a temperature of 700°C, and a 2 hr heating time produced waste forms with plutonium recoveries of 4±1 g/kg. With an increase in the frit to residue ratio, waste forms fabricated at this scale should meet the Rocky Flats product specification. The scale-up of the waste form fabrication process to nominally 3 kg is expected to require a 5:1 to 6:1 frit to residue ratio and maintaining the waste form centerline temperature at 700°C for 2 hr.

Demonstration of the defense waste processing facility vitrification process for tank 42 radioactive sludge – melter feed preparation

AbstractThe Defense Waste Processing Facility (DWPF), at the Savannah River Site (SRS), is processing and immobilizing the radioactive high level waste sludge at SRS into a durable borosilicate glass for final geological disposal. The DWPF is currently processing the second, million gallon batch of radioactive sludge. This second batch is primarily from Tank 42. Each time a new batch of radioactive sludge is to be processed by the DWPF, the process flowsheet is to be tested and demonstrated to ensure an acceptable melter feed and glass can be made. This demonstration was completed in the Shielded Cells Facility in the Savannah River Technology Center at SRS.This paper presents the processing and offgas data, and compositional analyses obtained during the preparation of a melter feed for this demonstration. A second paper in this conference describes the properties of the glass produced from this feed. The demonstration used Tank 42 sludge slurry and the DWPF process control strategy for blending the sludge slurry with Frit 200 to make an acceptable melter feed. To prepare feed for the melter, the flowsheet requires that the radioactive sludge slurry be treated with nitric and formic acid to adjust rheology and remove mercury. During this step, hydrogen is formed from the decomposition of the formic acid. The acidified sludge slurry is then mixed with the prescribed amount of glass forming frit and evaporated to the proper weight percent solids to prepare feed to the melter. During this step hydrogen is also formed. Results indicate that the H2 generation rate is below the DWPF safety limits and an acceptable melter feed was produced.

Field deployable tritium analysis system for ground and surface water measurements

Researchers from the Savannah River Technology Center, the Center for Applied Isotope Studies (CAIS) and Sampling Systems have developed a prototype Field Deployable Tritium Analysis System (FDTAS) for near-real-time measurements of environmental levels of tritium in ground and surface water. The device consists of a modified liquid scintillation counter coupled to an automatic sampler which incorporates on-line water purification. The FDTAS has been field tested at several Savannah River Site locations and has produced results comparable to laboratory analyses for low concentrations of tritium. Figures of merit obtained in the field include an average tritium background count rate of 1.5 counts per minute (cpm), tritium detection efficiency of ≈25%, and a detection limit of <10 Bq/l for a 100 minute count.

Determination of technetium-99 in aqueous samples by isotope dilution inductively coupled plasma-mass spectrometry

An isotope dilution/inductively coupled plasma mass spectrometric method (ID/ICP-MS) for measuring the concentration of technetium-99 in aqueous samples was developed at the Savannah River Technology Center (SRTC). The procedure is faster than radiometric tecniques, is less subject to interferences, and has equal or better detection limits. It is currently being used to measure the concentration of99Tc in samples of Savannah River water collected in the vicinity of the Savannah River Site. In this method one liter samples of water are spiked with97Tc. After equilibration, the technetium is extracted from the sample with a chromatographic resin. Interfering elements, molybdenum and ruthenium, are either not retained by the resin or are washed off with, dilute nitric acid. The technetium is then eluted with more concentrated nitric acid, and the99Tc/97Tc ratio in the eluant is measured with an ICP-MS. The99Tc concentration in the original sample is calculated from the99Tc/97Tc ratio. The chemical recovery of the extraction procedure is greater than 90%. The detection limit of the instrument, taken as three times the background counts atm/z=99, is 0.6 part per trillion (ppt). The detection limit of the procedure, taken as three times the standard deviation of several reagent blank analyses, is 0.33 pCi/l.

Startup and Operation of a Metal Hydride Based Isotope Separation Process

Production scale separation of tritium from other hydrogen isotopes at the Savannah River Site (SRS) in Aiken, SC, USA, has been accomplished by several methods. These methods include thermal diffusion (1957–1986), fractional absorption (1964–1968), and cryogenic distillation (1967–present). Most recently, the Thermal Cycling Absorption Process (TCAP), a metal hydride based hydrogen isotope separation system, began production in the Replacement Tritium Facility (RTF) on April 9, 1994. TCAP has been in development at the Savannah River Technology Center since 1980. The production startup of this semi-continuous gas chromatographic separation process is a significant accomplishment for the Savannah River Site and was achieved after years of design, development, and testing.

Consolidated Incineration Facility Tritium Emissions Monitoring

The Savannah River Technology Center, a research and development facility at the US Department of Energy`s Savannah River Site, provides environmental and regulatory compliance support to onsite operations. A new consolidated Incinerator Facility at SRS is being built to treat hazardous and a combination of hazardous and radioactive (mixed) wastes.

Laser ablation sampling with inductively coupled plasma atomic emission spectrometry for the analysis of prototypical glasses

Laser ablation sampling is presented as an alternative to dissolution procedures for elemental analyses of prototypical glasses using inductively coupled plasma atomic emission spectrometry. These glass samples were prototypes of vitrified waste products from the Savannah River Technology Center. The samples were not translated or rotated during laser sampling, but were repetitively sampled at a single spot using a KrF excimer laser with a 10 Hz repetition rate. The time-dependent mass ablation rate was measured and is discussed. Silicon, the major element in the matrix, was used as an internal standard, and excellent precision (sr= 1–3%) was obtained. Quantitative analysis was demonstrated using known prototypical glass compositions. Preferential vaporization was investigated by comparing measured elemental ratios using a nanosecond excimer laser (λ= 248 nm) and a picosecond Nd: YAG laser (fourth harmonic, λ= 266 nm).

Hydrogen Generation During Treatment of Simulated High-Level Radioactive Waste with Formic Acid

The Integrated Defense Waste Processing Facility (DWPF) Melter System (IDMS), operated by the Savannah River Technology Center, is a one-fifth scale pilot facility used in support of the startup and operation of the U.S. Department of Energy’s DWPF. Seven IDMS runs examined the effect of noble metals in simulated high-level radioactive waste (HLW) and important process variables on the generation of H2 during the preparation of melter feed with formic acid. The results showed that due to the noble metals in actual HLW, the lower flammable limit of H2 in air (4 vol%) could be exceeded in D WPF vessels, depending on such factors as off gas generation and air inleakage. A small but detectable quantity of H2 was generated even in the absence of noble metals. The results also verified that the most important process variable that affected the H2 generation rate was the amount of formic acid added to the system.Forced air purge systems with H2 monitoring instruments were installed in the DWPF to control the concentration of H2 in the offgas by fuel dilution during melter feed preparation. The design-basis forced air purge flow rate required in the DWPF during radioactive operations was based on the peak H2 generation rate observed during an IDMS run operated with 25% excess formic acid. This amount of excess formic acid was deemed a credible deviation from nominal operating conditions; therefore, a margin of safety was included in the design basis.