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Mechanical stress is an important modulator of lung morphogenesis, postnatal lung development, and compensatory lung regrowth. The effect of mechanical stress on stem or progenitor cells is unclear. We examined whether proliferative responses of epithelial progenitor cells, including dually immunoreactive (CCSP and proSP-C) progenitor cells (CCSP+/SP-C+) and type II alveolar epithelial cells (ATII), are affected by physical factors found in the lung of emphysematics, including loss of elastic recoil, reduced elastin content, and alveolar destruction. Mice underwent single lung pneumonectomy (PNY) to modulate transpulmonary pressure (mechanical stress) and to stimulate lung regeneration. Control mice underwent sham thoracotomy. Plombage of different levels was employed to partially or completely abolish this mechanical stress. Responses to graded changes in transpulmonary pressure were assessed in elastin-insufficient mice (elastin +/−, ELN+/−) and elastase-treated mice with elastase-induced emphysema. Physiological regrowth, morphometry (linear mean intercept; Lmi), and the proliferative responses of CCSP+/SP-C+, Clara cells, and ATII were evaluated. Plombage following PNY significantly reduced transpulmonary pressure, regrowth, and CCSP+/SP-C+, Clara cell, and ATII proliferation following PNY. In the ELN+/− group, CCSP+/SP-C+ and ATII proliferation responses were completely abolished, although compensatory lung regrowth was not significantly altered. In contrast, in elastase-injured mice, compensatory lung regrowth was significantly reduced, and ATII but not CCSP+/SP-C+ proliferation responses were impaired. Elastase injury also reduced the baseline abundance of CCSP+/SP-C+, and CCSP+/SP-C+ were found to be displaced from the bronchioalveolar duct junction. These data suggest that qualities of the extracellular matrix including elastin content, mechanical stress, and alveolar integrity strongly influence the regenerative capacity of the lung, and the patterns of cell proliferation in the lungs of adult mice.

The baseline flowsheet for low activity waste (LAW) in the Hanford River Protection Project (RPP) Waste Treatment Plant (WTP) includes pretreatment of supernatant by removing cesium using ion exchange. When the ion exchange column is loaded, the cesium will be eluted with a 0.5M nitric acid (HNO3) solution to allow the column to be conditioned for re-use. The cesium eluate solution will then be concentrated in a vacuum evaporator to minimize storage volume and to recycle HNO3.

The baseline flowsheet for low activity waste (LAW) in the Hanford River Protection Project (RPP) Waste Treatment Plant (WTP) includes pretreatment of supernatant by removing cesium using ion exchange. When the ion exchange column is loaded, the cesium will be eluted with a 0.5M nitric acid (HNO3) solution to allow the column to be conditioned for re-use. The cesium eluate solution will then be concentrated in a vacuum evaporator to minimize storage volume and recycle HNO3. To prevent the formation of solids during storage of the evaporator bottoms, criteria have been set for limiting the concentration of the evaporator product to 80 percent of saturation at 25 degrees C. Prior work has collected fundamental data for predicting solubility and other physical property measurements. Other ongoing efforts have involved the development of a computer model to predict solubility and physical properties during evaporation. Evaporation experiments were conducted with cesium eluate simulant generated from a pilot scale experiment in the Thermal Fluids Lab (TFL) at the Savannah River Technology Center (SRTC). The data from the experiments will be used to validate the modeling data.

Bechtel National Inc. (BNI) is using IBC Company's SuperLigand ion exchange resins to separate Cs and Tc from low-activity waste (LAW) solutions (IBC-1996). Cesium is removed using the SuperLig(R) 644 resin. The resin is then eluted after each use cycle with 0.5M nitric acid solution. BNI is planning to evaporate the Cs eluate solution to reduce the storage volume and recover eluate for re-use. The primary issue associated with evaporation is end point, or salt matrix solubility. To preclude formation of solids during the storage of evaporator products, an additional criteria has been set that limits the concentration of the evaporator bottoms to 80 percent of saturation at 25 degrees C. As a result, an understanding of the effects of constituent species on the bulk solubility must be developed prior to effective evaporator operations.

The Savannah River Site H-Canyon generates aqueous solutions that must be disposed of in the tank farm. Tests were conducted to observe the amount of TBP that is stripped as a function of volume of liquid stripped.

The Savannah River Site has enriched uranium (EU) solution which has been stored for almost 10 years since being purified in the second uranium cycle of the H area solvent extraction process. The preliminary SRTC data, in conjunction with information in the literature, is promising. However, very few experiments have been run, and none of the results have been confirmed with repeat tests. As a result, it is believed that insufficient data exists at this time to warrant Separations making any process or program changes based on the information contained in this report. When this data is confirmed in future testing, recommendations will be presented.

Experimental progress has been made in a wide range of areas which support the continued development of the nitric-phosphoric acid oxidation process for combustible, solid organic wastes. An improved understanding of the overall process operation has been obtained, acid recovery and recycle systems have been studied, safety issues have been addressed, two potential final waste forms have been tested, preliminary mass flow diagrams have been prepared, and process flowsheets have been developed. The flowsheet developed is essentially a closed-loop system which addresses all of the internally generated waste streams. The combined activities aim to provide the basis for building and testing a 250-400 liter pilot-scale unit. Variations of the process now must be evaluated in order to address the needs of the primary customer, SRS Solid Waste Management. The customer is interested in treating job control waste contaminated with Pu-238 for shipment to WIPP. As a result, variations for feed preparation, acid recycle, and final form manufacturing must be considered to provide for simpler processing to accommodate operations in high radiation and contamination environments. The purpose of this program is to demonstrate a nitric-phosphoric acid destruction technology which can treat a heterogeneous waste by oxidizing the solid and liquid organic compounds while decontaminating noncombustible items.

The Savannah River Site has enriched uranium (EU) solution which has been stored for almost 10 years since being purified in the second uranium cycle of the H area solvent extraction process. The concentrations in solution are approximately 6 g/L U and about 0.1 M nitric acid. Residual tributylphosphate in the solutions has slowly hydrolyzed to form dibutyl phosphoric acid (HDBP) at concentrations averaging 50 mg/L. Uranium is known to form compounds with the dibutylphosphate ion (DBP) which have limited solubility. The potential to form uranium-DBP solids raises a nuclear criticality safety issue. Prior SRTC tests (WSRC-TR-98-00188) showed that U-DBP solids precipitate at concentrations potentially attainable during the storage of enriched uranium solutions. Furthermore, evaporation of the existing EUS solution without additional acidification could result in the precipitation of U-DBP solids if the DBP concentration in the resulting solution exceeds 110 mg/L at ambient temperature. The same potential exists for evaporation of unwashed 1CU solutions. As a follow-up to the earlier studies, SRTC studied the solubility limits for solutions containing acid concentrations above 0.5M HNO3. The data obtained in these tests reveals a shift to higher levels of DBP solubility above 0.5M HNO3 for both 6 g/L and 12 g/L uranium solutions. Analysis of U-DBP solids from the tests identified a mixture of different molecular structures for the solids created. The analysis distinguished UO2(DBP)2 as the dominant compound present at low acid concentrations. As the acid concentration increases, the crystalline UO2(DBP)2 shows molecular substitutions and an increase in amorphous content. Further analysis by methods not available at SRS will be needed to better identify the specific compounds present. This data indicates that acidification prior to evaporation can be used to increase the margin of safety for the storage of the EUS solutions. Subsequent experimentation evaluated options for absorbing HDBP from solution using either activated carbon or anion exchange resin. The activated carbon outperformed the anion exchange resin. Activated carbon absorbs DBP rapidly and has demonstrated the capability of absorbing 15 mg of DBP per gram of activated carbon. Analytical results also show that activated carbon absorbs uranium up to 17 mg per gram of carbon. It is speculated that the uranium absorbed is part of a soluble U-DBP complex that has been absorbed. Additional testing must still be performed to 1) establish absorption limits for uranium for anion exchange resin, 2) evaluate desorption characteristics of uranium and DBP, and 3) study the possibility of re-using the absorbent.

The Savannah River Site (SRS) is working to dispose of the inventory of enriched uranium (EU) formerly used to make fuel for production reactors. The Tennessee Valley Authority (TVA) has agreed to take the material after blending the EU with either natural or depleted uranium to give a {sup 235}U concentration of 4.8 percent low-enriched uranium will be fabricated by a vendor into reactor fuel for use in TVA reactors. SRS prefers to blend the EU with existing depleted uranium (DU) solutions, however, the impurity concentrations in the DU and EU are so high that the blended material may not meet specifications agreed to with TVA. The principal non-radioactive impurities of concern are carbon, iron, phosphorus and sulfur. Neptunium and plutonium contamination levels are about 40 times greater than the desired specification. Tests of solvent extraction and fuel preparation with solutions of SRS uranium demonstrate that the UO{sub 2} prepared from these solutions will meet specifications for Fe, P and S, but may not meet the specifications for carbon. The reasons for carbon remaining in the oxide at such high levels is not fully understood, but may be overcome either by treatment of the solutions with activated carbon or heating the UO{sub 3} in air for a longer time during the calcination step of fuel preparation.Calculations of the expected removal of Np and Pu from the solutions show that the specification cannot be met with a single cycle of solvent extraction. The only way to ensure meeting the specification is dilution with natural U which contains no Np or Pu. Estimations of the decontamination from fission products and daughter products in the decay chains for the U isotopes show that the specification of 110 MEV Bq/g U can be met as long as the activities of the daughters of U- 235 and U-238 are excluded from the specification.

A general process is being developed for the treatment of solid TRU and hazardous organic waste. Experimental data indicates that 100 lb/hr of aliphatic organic (plastics) and 1,000 lb/hr of non-aliphatic organic compounds can be quantitatively oxidized in a 1,000 gallon reaction vessel. The process uses dilute nitric acid in a concentrated phosphoric acid media as the main oxidant for the organic compounds. Phosphoric acid allows oxidation at temperatures up to 200{degrees}C and is relatively non-corrosive on 304-L stainless steel, especially at room temperature. Many organic materials have been completely oxidized to CO{sub 2}, CO, and inorganic acids in a 0.1M HNO{sub 3}/14.8M H{sub 3}PO{sub 4} solution. Addition of 0.001M Pd{sup 2+} reduces the CO to near 1% of the released carbon gases. To accomplish complete oxidation the solution temperature must be maintained above 130--150{degrees}C. Organic materials quantitatively destroyed include neoprene, cellulose, EDTA, TBP, tartaric acid, and nitromethane. The oxidation is usually complete in a few hours for soluble organic materials. The oxidation rate for non-aliphatic organic solids is moderately fast and surface area dependent. Polyethylene is quantitatively oxidized in 1.0M HNO{sub 3}/13.8M H{sub 3}PO{sub 4} solution while contained in pressure vessels heated with microwave energy. This is probably duemore » to the high concentrations of NO{sub 2}{center_dot} obtained in the reaction environment.« less