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At least 145 records · Page 8

Cesium Ion Exchange Testing Using Crystalline Silicotitanate with Hanford Tank Waste 241-AP-107

At the time of this testing, the Low-Activity Waste Pretreatment System (LAWPS) was to provide for the initial production of immobilized low-activity waste by feeding Hanford tank supernate from tank farms to the Hanford Tank Waste Treatment and Immobilization Plant (WTP) Low-Activity Waste (LAW) Facility for immobilization. Washington River Protection Solutions requested that Hanford tank waste collected from tank 241-AP-107 (hereafter called AP-107) be processed using conceived pretreatment steps (suspended solids removal by filtration, Cs removal by ion exchange) then vitrified. A small-scale test platform to demonstrate the solids filtration, Cs removal, and LAW vitrification was constructed and installed at Pacific Northwest National Laboratory. Bench-scale ion exchange testing with approximately 9 L of AP-107 supernate was conducted using crystalline silicotitanate (CST) ion exchange media. The IONSIV R9140-B CST was provided by Honeywell UOP, LLC in 2018 (Batch 2081000057). The ion exchange media was first tested with simulant and was previously described. This report describes the Cs ion exchange batch contact and column test results with the AP-107 tank waste. Batch contact testing helps to evaluate CST performance on tank waste supernate prior to processing it in the ion exchange columns. Batch contacts were performed with the waste at four Cs concentrations at a phase ratio of 200 (liquid volume to exchanger mass) with AP-107. The distribution coefficient (K d ) at the equilibrium condition of 8.57 µg Cs/mL (AP-107 feed condition) was determined to be 669 mL AP-107/g CST. With a CST bed density of 1.00 g/mL, this K d corresponded to a predicted 50% Cs breakthrough of 669 bed volumes (BVs). The Cs load capacity at the equilibrium feed condition was determined to be 7.5 mg Cs/g dry CST. The column testing was prototypic to the intended LAWPS operations in a lead-lag column format, although on a small-scale basis with 10-mL CST beds. The feed was processed downflow through the lead column and then through the lag column at ~2.2 BV/h. Loading continued until the lag column reached the WTP waste acceptance criteria (WAC) for receiving supernatant waste for vitrification (a function of the Na and 137 Cs concentrations). For AP-107, the WAC is 0.114% of the influent 137 Cs concentration; this required a Cs decontamination factor of 876. The Cs effluent from the lag column reached the WAC after processing ~410 BVs. To keep the subsequent product effluent below the WAC, a replacement lag column was prepared, the lead column was removed from service (after processing a total of 471 BVs), the lag column was put into the lead column position, and the replacement lag column was installed. Feed processing continued and after another ~290 BVs the Cs effluent from the lag column again exceeded the WAC. In both cases, the lead columns only reached 25% Cs breakthrough before removal. Although 50% Cs breakthrough was not reached, this value was estimated and averaged based on extrapolation of the loading curves (640 BVs) and agreed within 4% of the predicted 50% Cs breakthrough from batch contact test results (669 BVs). Table ES.1 summarizes the observed column performance and relevant Cs loading characteristics.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Life cycle Management of 324 Project High-Activity, Mixed/Low-Level Radioactive Waste - 20506

U.S. DOE contractor CH2M Hill Plateau Remediation Company (CHPRC) manages the 324 Building and is preparing to remotely excavate and disposition the radioactive soil beneath the building to allow further deactivation of the building. The 324 Building is a non-reactor Category 2 Nuclear Facility located in the 300 Area of the Hanford Site. Records indicate that in October 1986, approximately 516 liters of a concentrated liquid waste stream containing cesium-137 ({sup 137}Cs) and strontium-90 ({sup 90}Sr) spilled onto the floor of a hot cell, B-Cell, in the 324 Building. The spill contained an estimated 1.3 million curies (Ci) of radioactivity. A breach in the sump of B-Cell was discovered in November 2009 during characterization of the soil under the building. Exposure rates in excess of 10,000 R/hr were detected through a system of access pipes installed under the hot cell, and an estimated 1.557 E+05 Ci of {sup 137}Cs and 6.842 E+04 Ci of {sup 90}Sr are in the soil beneath B-Cell. The magnitude of the soil contamination below B-Cell represents one of the most challenging remediation activities in the DOE complex. The objective of CHPRC's 324 Project is to remotely excavate and disposition the highly contaminated soil under the hot cell so building deactivation and stabilization activities can resume. Depending on the volume and radioactivity of contaminated soil under B-Cell, the material will be dispositioned either at the Hanford Site's Environmental Restoration Disposal Facility (ERDF) or in grout monoliths that will be created in the hot cells (A, C and D) next to B-Cell during future building demolition activities. The current project scope includes installing structural supports under the walls of B-Cell and using remotely operated equipment to remove debris from the hot cell, remove debris mixed with grout on the floor, saw through and remove the cell floor and liner, and excavate contaminated soil under B-Cell to a depth of up to 3.66 meters (12 feet). Addition or removal of debris in A-Cell may be required to support B-Cell cleanout and excavation activities. The base approach is to disposition debris and soil bins in the Radiochemical Engineering Cell (REC) monoliths, and subsequently disposition the monoliths during building demolition. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Tank 48H Tetraphenylborate Mitigation: Simulant Studies using Sodium Permanganate

Tank 48H currently holds legacy material containing organic tetraphenylborate (TPB) compounds from the operation of the In-Tank Precipitation (ITP) process. TPB was added during the ITP process to precipitate the otherwise soluble cesium as insoluble cesium TPB (CsTPB), but excessive benzene generation from TPB degradation curtailed this treatment method. The contents of Tank 48H, which include approximately 26.000 kg of potassium TPB (KTPB) and trace CsTPB, are not compatible with the waste treatment facilities at the Savannah River Site (SRS) since the organic content and the associated flammability issues pose a challenge to the salt processing and sludge processing facilities within the liquid waste system. An in-tank process to remove (or decompose) TPB safely would be of great value. Previous testing at Savannah River National Laboratory (SRNL) demonstrated the destruction of glycolate via chemical oxidation using sodium permanganate with simulated and radioactive waste. Scoping tests were performed to study the destruction of TPB to determine if the contents of Tank 48H would be amenable to the same type of destruction. Partial destruction of TPB was observed in Tank 48H simulants under mild conditions (e g., pH 11, room temperature) with no definitive indication of benzene generation. To build upon the success of the scoping tests, an additional study was requested to provide a better understanding of the underlying chemistry for Tank 48H content destruction using sodium permanganate. Three experiments were performed with Tank 48H simulants at 40 °C to determine the efficacy of using sodium permanganate for TPB destruction. Three starting pH values were selected: 1) pH 11 for comparison with the previous work at room temperature, 2) pH 10 as the minimum pH recommended by the Corrosion Control Program (CCP) for in-tank processing, and 3) pH 8 to determine the effectiveness of TPB destruction at near neutral pH. While below the allowable pH for the CCP, the experiment at pH 8 was performed to study the TPB-Permanganate reaction under more extreme conditions and further verify the potential for out-of-tank processing.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Synthesis and structural properties of a 2D Zn( II ) dodecahydroxy- closo -dodecaborate coordination polymer

Here, in this work, we discuss the synthesis and characterization of a 2D coordination polymer composed of a dianionic perhydroxylated boron cluster, [B 12 (OH) 12 2– ], coordinated to Zn( II )—the first example of a transition metal-coordinated [B 12 (OH) 12 ] 2– compound. This material was synthesized via cation exchange from the starting cesium salt and then subjected to rigorous characterization prior to and after thermal activation. Numerous techniques, including XRD, FTIR, SEM, TGA, and solid-state NMR revealed a 2D coordination polymer composed of sheets of Zn( II ) ions intercalated between planes of boron clusters. The as-synthesized material was then evacuated of solvent via thermal treatment, and atomic-level changes from this transformation were elucidated through a combination of 1D and 2D solid-state NMR analyses of 11 B and 1 H nuclei, suggesting the full removal of coordinated solvent molecules. Evidence also suggested that [B 12 (OH) 12 2– ] can adjust its coordination to Zn( II ) in the solid-state through hemilability of its numerous –OH ligands.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Colorimetric Detection and Surface Decontamination of Metal Ions by Using Strippable Polymers - 20024

Strippable coating is a low cost and efficient decontamination method far more compatible for the cleaning of surfaces. This method allows contaminant entrapment and removal from different types of surfaces and also characterized by low-secondary waste generation after decontamination. The present work deals with the elaboration and testing of a new composition of Polyvinyl alcohol (PVA) aqueous solutions utilize the TTA (Thenoyl trifluoroacetone) ligand, employed as a strippable coating for the removal of metal ions e.g., UO{sub 2}{sup +2}, Ni{sup +2}, Cu{sup +2}, Fe{sup +3}, V{sup +5}, Eu{sup +3} and Cs{sup +} from 'contaminated' surfaces. The study aims at developing contaminant-sensing and decontaminating coatings for the detection and removal of metal ions from non porous surfaces. This 'smart' coating is shown to exhibit a clear contaminant dependent response in the form of color variation, and a detection sensitivity less than 0.5 μg per cm{sup 2}. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Efficient Mn 2+ Doping in Non-Stoichiometric Cesium Lead Bromide Perovskite Quantum Dots

Doping magnetic transition metal ions (e.g., Mn 2+ ) into colloidal quantum dots endows novel optical and magnetic properties to the host materials. CsPbBr 3 quantum dots (QDs) are emerging light-emitting materials with high structural and chemical flexibility in the visible spectral regime. However, efficiently doping Mn 2+ ions in CsPbBr 3 QDs remains challenging, especially when size confinement and ensemble uniformity are needed for understanding the underexplored exciton-dopant exchange interaction. Here, we introduce a doping mechanism based on electrostatic surface Mn 2+ adsorption that enables efficient Mn 2+ incorporation in strongly confined CsPbBr 3 QDs. The resultant QDs are found to have a Cs-deficient stoichiometry compared to their undoped counterparts. A redox reaction-based purification method was developed to remove Mn 2+ cations that are tightly adsorbed on the surface to determine the concentration of lattice-incorporated Mn 2+ . Our synthesis enables a Mn 2+ doping/alloying concentration of up to ∼44% with a Mn 2+ photoluminescence efficiency exceeding 90%. This allows for the determination of the intrinsic exciton-to-dopant energy transfer rate.

Hidayatova, Lamia [Univ. of Oklahoma, Norman, OK (↗

Room temperature doping of Ln 3+ in perovskite nanoparticles: a halide exchange mediated cation exchange approach

This study presents a halide exchange mediated cation exchange strategy for a room temperature doping of trivalent lanthanide cations (Ln 3+ ) in cesium lead halide (CsPbX 3 ) nanoparticles (NPs). Post-synthetic addition of LnCl 3 [Ln = Nd, Sm, Eu, Tb, Dy, and Yb] to a solution of CsPbBr 3 NPs generates the corresponding lanthanide doped NPs which display host sensitized Ln 3+ emission. Structural and spectroscopic characterizations indicate a successful halide exchange and substitutional displacement of Pb 2+ by Ln 3+ . The effect of halide identity in controlling the Ln 3+ sensitization was also evaluated. Here, a photophysical framework is presented that can be used to predict the Ln 3+ sensitization in perovskite NPs semiempirically, thereby removing the constraints of trial and error in designing a perovskite NP-Ln 3+ host–guest combination.

77 NANOSCIENCE AND NANOTECHNOLOGY↗

A target for production of radioxenons

A liquid cesium target has been developed which permits the production and identification of the neutron deficient isotopes of xenon. The thick target yields of Xe-127 and Xe-129 m were measured for proton energies between 28 and 40 MeV. Production of Xe-129 m, as indicated by the presence of a 196 keV line in the gamma ray spectra, occurred at all proton energies; Xe-127 was produced at proton energies above 33 MeV. Both accumulated on the target with the same time constant, and the cross sections were comparable in magnitude. The radioxenons were removed from the target by a stream of carrier gas so that chemical separation was unnecessary.

Blue, J. W.↗

Influence of Cs Loading on Pt/m-ZrO 2 Water–Gas Shift Catalysts

Certain alkali metals (Na, K) at targeted loadings have been shown in recent decades to significantly promote the LT-WGS reaction. This occurs at alkali doping levels where a redshift in the C-H band of formate occurs, indicating electronic weakening of the bond. The C-H bond breaking of formate is the proposed rate-limiting step of the formate associative mechanism, lending support to the occurrence of this mechanism in H2-rich environments of the LT-WGS stage of fuel processors. Continuing in this vein of research, 2%Pt/m-ZrO 2 was promoted with various levels of Cs in order to explore its influence on the rate of formate intermediate decomposition, as well as that of LT-WGS in a fixed bed reactor. In situ DRIFTS experiments revealed that Cs promoter loadings of 3.87% to 7.22% resulted in significant acceleration of the forward formate decomposition in steam at 130 °C. Of all of the alkali metals tested to date, the redshift in the formate ν(CH) band with the incorporation of Cs was the greatest. XANES difference experiments at the Pt L 2 and L 3 edges indicated that the electronic effect was not likely due to an enrichment of electronic density on Pt. CO 2 TPD experiments revealed that, unlike Na and K promoters, Cs behaves more like Rb in that the decomposition of the second intermediate in LT-WGS, carbonate species, is hindered due to (1) increased basicity of Cs, (2) the tendency of Cs to cover Pt sites that facilitate CO 2 decomposition, and (3) the tendency of Cs to increase Pt particle size as shown by EXAFS results, resulting in fewer Pt sites that facilitate CO2 decomposition. As such, the LT-WGS rate was hindered overall and the rate-limiting step shifted to carbonate decomposition (CO 2 removal). Like its Rb counterpart, low levels of added Cs (e.g., 0.72%Cs) were found to improve the stability of the catalyst relative to the unpromoted catalyst; the stability comparison was made at similar CO conversion level as well as similar space velocity.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

NaK Variable Conductance Heat Pipe for Radioisotope Stirling Systems

In a Stirling radioisotope power system, heat must continually be removed from the General Purpose Heat Source (GPHS) modules to maintain the modules and surrounding insulation at acceptable temperatures. The Stirling convertor normally provides most of this cooling. If the Stirling convertor stops in the current system, the insulation is designed to spoil, preventing damage to the GPHS, but also ending use of that convertor for the mission. An alkali-metal Variable Conductance Heat Pipe (VCHP) was designed to allow multiple stops and restarts of the Stirling convertor. In the design of the VCHP for the Advanced Stirling Radioisotope Generator, the VCHP reservoir temperature can vary between 40 and 120 C. While sodium, potassium, or cesium could be used as the working fluid, their melting temperatures are above the minimum reservoir temperature, allowing working fluid to freeze in the reservoir. In contrast, the melting point of NaK is -12 C, so NaK can't freeze in the reservoir. One potential problem with NaK as a working fluid is that previous tests with NaK heat pipes have shown that NaK heat pipes can develop temperature non-uniformities in the evaporator due to NaK's binary composition. A NaK heat pipe was fabricated to measure the temperature non-uniformities in a scale model of the VCHP for the Stirling Radioisotope system. The temperature profiles in the evaporator and condenser were measured as a function of operating temperature and power. The largest delta T across the condenser was 2S C. However, the condenser delta T decreased to 16 C for the 775 C vapor temperature at the highest heat flux applied, 7.21 W/ square cm. This decrease with increasing heat flux was caused by the increased mixing of the sodium and potassium in the vapor. This temperature differential is similar to the temperature variation in this ASRG heat transfer interface without a heat pipe, so NaK can be used as the VCHP working fluid.

Tarau, Calin↗

Tailoring Crystallization Dynamics of CsPbI 3 for Scalable Production of Efficient Inorganic Perovskite Solar Cells

All-inorganic perovskite cesium lead triiodide (CsPbI 3 ) with inorganic nature, low-temperature synthesis, and a suitable bandgap is desirable for high-performance photovoltaics. However, the scalable production of CsPbI 3 photovoltaics is still challenging due to a large nucleation energy barrier and slow phase transition during unassisted natural crystallization. Here, the crystallization dynamics of CsPbI 3 thin films is tailored via lead acetate (PbAc 2 ) substitution in the perovskite precursor ink, allowing the scalable fabrication of efficient all-inorganic perovskite solar cells and minimodules. Introducing PbAc 2 enlarges CsPbI 3 colloid size in the precursor and reduces the nucleation energy barrier. Additionally, reactions between acetate and dimethylammonium in the wet film accelerate the removal of dimethylammonium additives and generate solvent vapors for self-regulate internal solvent annealing, resulting in densely packed, uniform, and pinhole-free CsPbI 3 perovskite films over large areas. This strategy demonstrates inverted CsPbI 3 solar cells with 20.17% efficiency and good operational stability (retaining 95.5% of initial efficiency after continuous operation for 1800 h) and 15.1%-efficient CsPbI 3 minimodules with an active area of 26.8 cm 2 .

36 MATERIALS SCIENCE↗

Cesium intercalation of graphene: A 2D protective layer on alkali antimonide photocathode

Alkali antimonide photocathodes have wide applications in free-electron lasers and electron cooling. The short lifetime of alkali antimonide photocathodes necessitates frequent replacement of the photocathodes during a beam operation. Furthermore, exposure to mediocre vacuum causes loss of photocathode quantum efficiency due to the chemical reaction with residual gas molecules. Theoretical analyses have shown that covering an alkali antimonide photocathode with a monolayer graphene or hexagonal boron nitride protects it in a coarse vacuum environment due to the inhibition of chemical reactions with residual gas molecules. Alkali antimonide photocathodes require an ultra-high vacuum environment, and depositing a monolayer 2D material on it poses a serious challenge. In the present work, we have incorporated a novel method known as intercalation, in which alkali atoms pass through the defects of a graphene thin film to create a photocathode material underneath. Initially, Sb was deposited on a Si substrate, and a monolayer graphene was transferred on top of the Sb film. Heat cleaning around 550–600 °C effectively removed the Sb oxides, leaving metallic Sb underneath the graphene layer. Depositing Cs on top of a monolayer graphene enabled the intercalation process. Atomic force microscopy, Raman spectroscopy, x-ray photoelectron spectroscopy, low energy electron microscopy, and x-ray diffraction measurements were performed to evaluate photocathode formation underneath the monolayer graphene. Our analysis shows that Cs penetrated the graphene and reacted with Sb and formed Cs 3 Sb.

2D materials↗

The Influence of Alkalinity on the Uptake of Cs{sup +} and Sr{sup 2+} by Cation-substituted Natisites in Sodium-bearing Conditions - 20336

The titanosilicate natisite (Na{sub 2}TiSiO{sub 5}) is a kinetic phase of the mineral sitinakite (Na{sub 2}Ti{sub 2}O{sub 3}SiO{sub 4}.2H{sub 2}O), which is a reference material in the removal of Cs and Sr from radioactive high-level waste. Natisite is disregarded in the literature as a candidate sorbent for Cs and Sr, despite being more thermally stable than sitinakite, which is a critical property for this application. Replacing portion of the Ti in natisite by other metals is believed to enhance natisite sorption properties. In nuclear waste remediation, Cs and Sr are contained in high-salinity liquid wastes that can either be highly acidic or alkaline. In the present study, Al-, Sn-, and Zr-natisites were synthesized, and compared to pure natisite and sitinakite in batch experiments. Five concentrations of NaOH and NaNO{sub 3} (i.e. source of Na{sup +}) mimicking conditions of high alkalinity, and competing Na{sup +} ions, respectively, were evaluated. Sorption results demonstrate that sitinakite is generally more effective than all four natisites in removing Cs and Sr. However, sitinakite uptake mechanism seems to deteriorate at increasing concentrations of base and Na{sup +}. Although less selective than sitinakite, all natisite sorbents provided a less dramatic decline in Cs and Sr uptake through increasing molarities. Overall, there was an improvement in the sorption of Cs and Sr by the metal-substituted materials over natisite and, in neither of the testing solutions, natisite outperformed its substituted variants. In highly alkaline solutions, Sn- and Zr-Natisite provided for exceptional removal for Sr, removing more than twice the amount sorbed by sitinakite at 0.1 M NaOH. The results obtained for Sr uptake by all four natisite materials in alkaline solutions are promising, especially considering the highly alkaline nature of wastes from the nuclear industry. Future studies should investigate the sorption mechanisms responsible for natisite selectivity for Sr at high pH. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Penetration of Fission Products Ions into Complex Solids and the Effect of Ionic Wash Methods

During washing of radiologically impacted building surfaces, penetration of radionuclide ions into complex solids associated with these surfaces may occur. This study investigates the penetration of Cs-137, Sr-85, and Eu-152 solutions into numerous common building materials and radionuclide behavior when these materials were exposed to a static bath or low-pressure flow of tap water, 0.1 M potassium chloride (KCl), and 0.5 M KCl. The decontamination efficacy and the depth profile for residual contamination were measured to determine the conditions under which applying a wash solution has benefit compared to physically removing the surface material. On asphalt, 70-80% of the radionuclides were found to be within 0.02 mm of the surface. Concrete is more porous than asphalt, and 80% of the radionuclides were within 0.2 mm of the surface for Cs-137 and Eu-152 and 50-80% for Sr-85. Water effectively removed all contaminants from hard nonporous surfaces. Finally, this paper illustrates that a wash penalty factor concept-defined as ratio of the depth at which 50% of the radioactivity is found in the washed sample divided by the depth at which 50% of radioactivity is found in the control-can serve as a way to quantify whether the wash method increases the depth at which contamination penetrates into the material and thus the material becomes more difficult to decontaminate.

building materials↗

Graphene as reusable substrate for bialkali photocathodes

Bialkali photocathodes, such as cesium potassium antimonide (CsK 2 Sb), can generate a high-brightness electron beam using a high-power green laser. These photocathode materials have potential applications in advanced accelerators and electron microscopes. It is known that the quantum efficiency (QE) of these photocathodes is affected severely by their substrates; however, reusability of the substrates is not well known. Here, we use graphene, silicon (Si), and molybdenum (Mo) substrates to evaluate the effects of substrates on the QE of redeposited CsK 2 Sb photocathodes after thermal cleanings. We found the QE of CsK 2 Sb photocathodes redeposited on a graphene substrate after a thermal cleaning at 500 °C remained largely unchanged. On the other hand, the QE of redeposited photocathodes on Si and Mo substrates after thermal cleaning at the same temperature decreased drastically. We used X-ray photoelectron spectroscopy (XPS) to quantitatively evaluate the residues of photocathodes after thermal cleaning at 400 °C and 500 °C. We found that the Sb, K, and Cs are removed by thermal cleaning at 500 °C for the graphene substrate, but all or the majority of these elements remained on the Si and Mo substrates. The results were consistent with our density functional theory (DFT) calculations for the case of Si, which we investigated. Furthermore, our angle-resolved photoemission spectroscopy (ARPES) on graphene indicated that its intrinsic electronic structure is preserved after photocathode deposition and thermal cleaning at 500 °C. Hence, we attributed the difference in amount of photocathode residue to the unique dangling-bond- free surface of inert graphene. Our results provide a foundation for graphene-based reusable substrates for high-QE semiconductor photocathodes.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗

Decontamination of urban surfaces contaminated with radioactive materials and consequent onsite recycling of the waste water

Enhancing rapid remediation strategies is paramount for recovery after a large-scale nuclear contamination event in an urban environment. Some current strategies recommend use of readily available equipment, materials, and facilities to expedite recovery. For example, applying pressurized water to contaminated surfaces may effectively remove radioactive contamination. In this study, a commercial power washer removes soluble forms of 152 Eu 3 + , 85 Sr 2 + , and 137 Cs + contamination from common porous building materials, and computer simulations characterize the recycling of the resultant contaminated wash water. Pressure washing the porous building materials under spray conditions typical with do-it-yourself units improved decontamination factors (DFs) for 152 Eu compared to low-pressure application of tap water (majority of two-tailed t-test p-values < 0.1), but pressure did not improve DFs for 137 Cs or 85 Sr. For both pressurized and low-pressure applications, adding potassium ions (K + ) to promote ion exchange reactions produced significantly higher DFs for tested radionuclides on asphalt, brick, and concrete. The resultant contaminated wash water can be processed through self-prepared chemical filtration beds of clay and sand. Modeled in a prior study, the beds yielded linear trends (R 2 > 0.98) in sensitivity analyses between most bed configuration variables and bed performance variables, permitting flexible ad-hoc bed design. The experimental and simulation results led to estimates of the remediation rate and waste generated after cleaning 250 m 2 of cesium-contaminated concrete from the combined deployment of a power washer and two different mobile treatment beds. Furthermore, the first treatment bed was designed to reduce treatment time and processed 1900 L of wash solution in 70 min using 880 kg of clay/sand infill material. Designed to reduce the solid waste generated, the second bed processed the same solution volume in 1040 min (17 h) using 170 kg of clay/sand infill material. The results of this analysis warrant further investigation of power washing with recycled salt solution as an effective rapid decontamination method with manageable waste.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

A laser-cooled cesium fountain frequency standard and a measurement of the frequency shift due to ultra-cold collisions

A frequency standard based on an atomic fountain of cesium atoms may have an accuracy of 10(exp -16) due to longer interaction times and smaller anticipated systematic errors. All of the known systematic effects that now limit the accuracy of the Cs frequency standard increase either linearly or as some higher power of the atom's velocity. The one systematic frequency shift which is dramatically different is the frequency shift due to the collisions between the laser cooled atoms. At a temperature of a few micro-K, the de Broglie wavelength (lambda(sub deB) = h/p, where h is Planck's constant and p is the momentum of the atom) is much larger than the scale of the interatomic potential. Under these conditions the collision cross sections can be as large as (lambda(sub deB)(sup 2))/Pi and the frequency shift due to these collisions was recently calculated. In our Cs atomic fountain, we laser cooled and trapped 10(exp 10) Cs atoms in 0.4 s. By shifting the frequencies of the laser beams, the atoms were launched upwards at 2.5 m/s and a fraction of the atoms were optically pumped into the F=3 ground state. The unwanted atoms in the F=4 ground state were removed from the fountain with radiation pressure from a laser beam tuned to excite only those atoms. The Cs atoms in the F=3 state traveled ballistically upwards, were excited by the microwave cavity, and then returned back through the same cavity in the atomic fountain configuration. By varying the cold atom density, a density dependent shift of -12.9 +/- 0.7 mHz or -1.4 x 10-12 for an average fountain density of (2.7 +/- 1.5) 10(exp 9) atoms/cm(sup 3) was measured.

Gibble, Kurt↗

Analysis in Support of Disposition of Tank 48 Legacy Material

This report contains the characterization of six 200-mL Tank 48H samples: HTF-48-21-74, HTF-48-21-75, HTF-48-21-76, HTF-48-21-81, HTF-48-21-82, and HTF-48-21-83. The effort supports a Systems Engineering Evaluation (SEE) recommendation involving a Tank 48H decantation strategy that would remove liquid volume and grout the solids. The first three Tank 48H samples were surface samples taken after a quiescent period in the tank. The quiescent period allowed settling of the solids, these being mostly potassium tetraphenylborate. These three surface samples had no measurable solids, though a settling haze could be seen. The latter three samples were taken at 48, 25, and 10 inches from the bottom of Tank 48H immediately after tank mixing pumps had been run. Those samples contained measurable insoluble solids that were readily visible. All six samples were analyzed to provide chemical and radionuclide concentrations as defined as the "Limit" and "Target" in the Saltstone Production Facility (SPF) Waste Acceptance Criteria (WAC) and per the compliance strategy in the Tank Farm Waste Compliance Plan (WCP). Samples were analyzed by many methods to determine pH, density/specific gravity, radioactive isotopes, soluble and insoluble elements, total solids, total insoluble solids, organic and inorganic mercury, volatile and semi-volatile chemicals, and anions. Photographs of the settling of small samples were taken over time and are displayed in this report. The extent of settling was very significant, showing that surface sample liquids are similar to filtrates. However, Cs-137 measurements exceeded WAC limits in all surface samples (1.1E+07 vs. 1.3E+06). The solids were found to contain very high cesium activity, measured as high as 1.1E+10 dpm/gram. Sodium was in the range of 4.24 to 4.74 M for all samples, so the solids would tend to settle with time and would not be at risk of floating without air entrainment. No organic mercury was detected in this work.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗