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Sample records for renal salt wasting

  1. Central Diabetes Insipidus and Cisplatin-Induced Renal Salt Wasting Syndrome: A Challenging Combination.

    Science.gov (United States)

    Cortina, Gerard; Hansford, Jordan R; Duke, Trevor

    2016-05-01

    We describe a 2-year-old female with a suprasellar primitive neuroectodermal tumor and central diabetes insipidus (DI) who developed polyuria with natriuresis and subsequent hyponatremia 36 hr after cisplatin administration. The marked urinary losses of sodium in combination with a negative sodium balance led to the diagnosis of cisplatin-induced renal salt wasting syndrome (RSWS). The subsequent clinical management is very challenging. Four weeks later she was discharged from ICU without neurological sequela. The combination of cisplatin-induced RSWS with DI can be confusing and needs careful clinical assessment as inaccurate diagnosis and management can result in increased neurological injury. © 2016 Wiley Periodicals, Inc.

  2. Prenatal programming of renal salt wasting resets postnatal salt appetite, which drives food intake in the rat.

    Science.gov (United States)

    Alwasel, Saleh H; Barker, David J P; Ashton, Nick

    2012-03-01

    Sodium retention has been proposed as the cause of hypertension in the LP rat (offspring exposed to a maternal low-protein diet in utero) model of developmental programming because of increased renal NKCC2 (Na+/K+/2Cl- co-transporter 2) expression. However, we have shown that LP rats excrete more rather than less sodium than controls, leading us to hypothesize that LP rats ingest more salt in order to maintain sodium balance. Rats were fed on either a 9% (low) or 18% (control) protein diet during pregnancy; male and female offspring were studied at 4 weeks of age. LP rats of both sexes held in metabolism cages excreted more sodium and urine than controls. When given water to drink, LP rats drank more and ate more food than controls, hence sodium intake matched excretion. However, when given a choice between saline and water to drink, the total volume of fluid ingested by LP rats fell to control levels, but the volume of saline taken was significantly larger [3.8±0.1 compared with 8.8±1.3 ml/24 h per 100 g of body weight in control and LP rats respectively; Psodium content and ECF (extracellular fluid) volumes were greater in LP rats. These results show that prenatal programming of renal sodium wasting leads to a compensatory increase in salt appetite in LP rats. We speculate that the need to maintain salt homoeostasis following malnutrition in utero stimulates greater food intake, leading to accelerated growth and raised BP (blood pressure).

  3. Cisplatin-Induced Renal Salt Wasting Requiring over 12 Liters of 3% Saline Replacement

    Directory of Open Access Journals (Sweden)

    Phuong-Chi Pham

    2017-01-01

    Full Text Available Cisplatin is known to induce Fanconi syndrome and renal salt wasting (RSW. RSW typically only requires transient normal saline (NS support. We report a severe RSW case that required 12 liters of 3% saline. A 57-year-old woman with limited stage small cell cancer was admitted for cisplatin (80 mg/m2 and etoposide (100 mg/m2 therapy. Patient’s serum sodium (SNa decreased from 138 to 133 and 125 mEq/L within 24 and 48 hours of cisplatin therapy, respectively. A diagnosis of syndrome of inappropriate antidiuretic hormone secretion (SIADH was initially made. Despite free water restriction, patient’s SNa continued to decrease in association with acute onset of headaches, nausea, and dizziness. Three percent saline (3%S infusion with rates up to 1400 mL/day was required to correct and maintain SNa at 135 mEq/L. Studies to evaluate Fanconi syndrome revealed hypophosphatemia and glucosuria in the absence of serum hyperglycemia. The natriuresis slowed down by 2.5 weeks, but 3%S support was continued for a total volume of 12 liters over 3.5 weeks. Attempts of questionable benefits to slow down glomerular filtration included the administration of ibuprofen and benazepril. To our knowledge, this is the most severe case of RSW ever reported with cisplatin.

  4. Solid waste disposal into salt mines

    International Nuclear Information System (INIS)

    Repke, W.

    1981-01-01

    The subject is discussed as follows: general introduction to disposal of radioactive waste; handling of solid nuclear waste; technology of final disposal, with specific reference to salt domes; conditioning of radioactive waste; safety barriers for radioactive waste; practice of final disposal in other countries. (U.K.)

  5. Waste treatment using molten salt oxidation

    International Nuclear Information System (INIS)

    Navratil, J.D.; Stewart, A.E.

    1996-01-01

    MSO technology can be characterized as a submerged oxidation process; the basic concept is to introduce air and wastes into a bed of molten salt, oxidize the organic wastes in the molten salt, use the heat of oxidation to keep the salt molten and remove the salt for disposal or processing and recycling. The molten salt (usually sodium carbonate at 900-1000 C) provides four waste management functions: providing a heat transfer medium, catalyzing the oxidation reaction, preventing the formation of acid gases by forming stable salts, and efficiently capturing ash particles and radioactive materials by the combined effects of wetting, encapsulation and dissolution. The MSO process requires no wet scrubbing system for off-gas treatment. The process has been developed through bench-scale and pilot-scale testing, with successful destruction demonstration of a wide variety of hazardous and mixed (radioactive and hazardous wastes). (author). 24 refs, 2 tabs, 2 figs

  6. Organic waste processing using molten salt oxidation

    Energy Technology Data Exchange (ETDEWEB)

    Adamson, M. G., LLNL

    1998-03-01

    Molten Salt Oxidation (MSO) is a thermal means of oxidizing (destroying) the organic constituents of mixed wastes, hazardous wastes, and energetic materials while retaining inorganic and radioactive constituents in the salt. For this reason, MSO is considered a promising alternative to incineration for the treatment of a variety of organic wastes. The U. S. Department of Energy`s Office of Environmental Management (DOE/EM) is currently funding research that will identify alternatives to incineration for the treatment of organic-based mixed wastes. (Mixed wastes are defined as waste streams which have both hazardous and radioactive properties.) One such project is Lawrence Livermore National Laboratory`s Expedited Technology Demonstration of Molten Salt Oxidation (MSO). The goal of this project is to conduct an integrated demonstration of MSO, including off-gas and spent salt treatment, and the preparation of robust solid final forms. Livermore National Laboratory (LLNL) has constructed an integrated pilot-scale MSO treatment system in which tests and demonstrations are presently being performed under carefully controlled (experimental) conditions. The system consists of a MSO process vessel with dedicated off-gas treatment, a salt recycle system, feed preparation equipment, and equipment for preparing ceramic final waste forms. In this paper we describe the integrated system and discuss its capabilities as well as preliminary process demonstration data. A primary purpose of these demonstrations is to identify the most suitable waste streams and waste types for MSO treatment.

  7. Permanent Disposal of Nuclear Waste in Salt

    Science.gov (United States)

    Hansen, F. D.

    2016-12-01

    Salt formations hold promise for eternal removal of nuclear waste from our biosphere. Germany and the United States have ample salt formations for this purpose, ranging from flat-bedded formations to geologically mature dome structures. Both nations are revisiting nuclear waste disposal options, accompanied by extensive collaboration on applied salt repository research, design, and operation. Salt formations provide isolation while geotechnical barriers reestablish impermeability after waste is placed in the geology. Between excavation and closure, physical, mechanical, thermal, chemical, and hydrological processes ensue. Salt response over a range of stress and temperature has been characterized for decades. Research practices employ refined test techniques and controls, which improve parameter assessment for features of the constitutive models. Extraordinary computational capabilities require exacting understanding of laboratory measurements and objective interpretation of modeling results. A repository for heat-generative nuclear waste provides an engineering challenge beyond common experience. Long-term evolution of the underground setting is precluded from direct observation or measurement. Therefore, analogues and modeling predictions are necessary to establish enduring safety functions. A strong case for granular salt reconsolidation and a focused research agenda support salt repository concepts that include safety-by-design. Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. Department of Energy's National Nuclear Security Administration under contract DE-AC04-94AL85000. Author: F. D. Hansen, Sandia National Laboratories

  8. Designing Advanced Ceramic Waste Forms for Electrochemical Processing Salt Waste

    International Nuclear Information System (INIS)

    Ebert, W. L.; Snyder, C. T.; Frank, Steven; Riley, Brian

    2016-01-01

    This report describes the scientific basis underlying the approach being followed to design and develop ''advanced'' glass-bonded sodalite ceramic waste form (ACWF) materials that can (1) accommodate higher salt waste loadings than the waste form developed in the 1990s for EBR-II waste salt and (2) provide greater flexibility for immobilizing extreme waste salt compositions. This is accomplished by using a binder glass having a much higher Na_2O content than glass compositions used previously to provide enough Na+ to react with all of the Cl- in the waste salt and generate the maximum amount of sodalite. The phase compositions and degradation behaviors of prototype ACWF products that were made using five new binder glass formulations and with 11-14 mass% representative LiCl/KCl-based salt waste were evaluated and compared with results of similar tests run with CWF products made using the original binder glass with 8 mass% of the same salt to demonstrate the approach and select a composition for further studies. About twice the amount of sodalite was generated in all ACWF materials and the microstructures and degradation behaviors confirmed our understanding of the reactions occurring during waste form production and the efficacy of the approach. However, the porosities of the resulting ACWF materials were higher than is desired. These results indicate the capacity of these ACWF waste forms to accommodate LiCl/KCl-based salt wastes becomes limited by porosity due to the low glass-to-sodalite volume ratio. Three of the new binder glass compositions were acceptable and there is no benefit to further increasing the Na content as initially planned. Instead, further studies are needed to develop and evaluate alternative production methods to decrease the porosity, such as by increasing the amount of binder glass in the formulation or by processing waste forms in a hot isostatic press. Increasing the amount of binder glass to eliminate porosity will decrease the waste

  9. Designing Advanced Ceramic Waste Forms for Electrochemical Processing Salt Waste

    Energy Technology Data Exchange (ETDEWEB)

    Ebert, W. L. [Argonne National Lab. (ANL), Argonne, IL (United States); Snyder, C. T. [Argonne National Lab. (ANL), Argonne, IL (United States); Frank, Steven [Argonne National Lab. (ANL), Argonne, IL (United States); Riley, Brian [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-03-01

    This report describes the scientific basis underlying the approach being followed to design and develop “advanced” glass-bonded sodalite ceramic waste form (ACWF) materials that can (1) accommodate higher salt waste loadings than the waste form developed in the 1990s for EBR-II waste salt and (2) provide greater flexibility for immobilizing extreme waste salt compositions. This is accomplished by using a binder glass having a much higher Na2O content than glass compositions used previously to provide enough Na+ to react with all of the Cl– in the waste salt and generate the maximum amount of sodalite. The phase compositions and degradation behaviors of prototype ACWF products that were made using five new binder glass formulations and with 11-14 mass% representative LiCl/KCl-based salt waste were evaluated and compared with results of similar tests run with CWF products made using the original binder glass with 8 mass% of the same salt to demonstrate the approach and select a composition for further studies. About twice the amount of sodalite was generated in all ACWF materials and the microstructures and degradation behaviors confirmed our understanding of the reactions occurring during waste form production and the efficacy of the approach. However, the porosities of the resulting ACWF materials were higher than is desired. These results indicate the capacity of these ACWF waste forms to accommodate LiCl/KCl-based salt wastes becomes limited by porosity due to the low glass-to-sodalite volume ratio. Three of the new binder glass compositions were acceptable and there is no benefit to further increasing the Na content as initially planned. Instead, further studies are needed to develop and evaluate alternative production methods to decrease the porosity, such as by increasing the amount of binder glass in the formulation or by processing waste forms in a hot isostatic press. Increasing the amount of binder glass to eliminate porosity will decrease

  10. Disposal of Savannah River Plant waste salt

    International Nuclear Information System (INIS)

    Dukes, M.D.

    1982-01-01

    Approximately 26-million gallons of soluble low-level waste salts will be produced during solidification of 6-million gallons of high-level defense waste in the proposed Defense Waste Processing Facility (DWPF) at the Savannah River Plant (SRP). Soluble wastes (primarily NaNO 3 , NaNO 2 , and NaOH) stored in the waste tanks will be decontaminated by ion exchange and solidified in concrete. The resulting salt-concrete mixture, saltcrete, will be placed in a landfill on the plantsite such that all applicable federal and state disposal criteria are met. Proposed NRC guidelines for the disposal of waste with the radionuclide content of SRP salt would permit shallow land burial. Federal and state rules require that potentially hazardous chemical wastes (mainly nitrate-nitrate salts in the saltcrete) be contained to the degree necessary to meet drinking water standards in the ground water beneath the landfill boundary. This paper describes the proposed saltcrete landfill and tests under way to ensure that the landfill will meet these criteria. The work includes laboratory and field tests of the saltcrete itself, a field test of a one-tenth linear scale model of the entire landfill system, and a numerical model of the system

  11. Immobilization of IFR salt wastes in mortar

    International Nuclear Information System (INIS)

    Fischer, D.F.; Johnson, T.R.

    1988-01-01

    Portland cement-base mortars are being considered for immobilizing chloride salt wastes produced by the fuel cycles of Integral Fast Reactors (IFR). The IFR is a sodium-cooled fast reactor with metal alloy fuels. It has a close-coupled fuel cycle in which fission products are separated from the actinides in an electrochemical cell operating at 500/degree/C. This cell has a liquid cadmium anode in which the fuels are dissolved and a liquid salt electrolyte. The salt will be a mixture of either lithium, potassium, and sodium chlorides or lithium, calcium, barium, and sodium chlorides. One method being considered for immobilizing the treated nontransuranic salt waste is to disperse the salt in a portland cement-base mortar that will be sealed in corrosion-resistant containers. For this application, the grout must be sufficiently fluid that it can be pumped into canister-molds where it will solidify into a strong, leach-resistant material. The set times must be longer than a few hours to allow sufficient time for processing, and the mortar must reach a reasonable compressive strength (/approximately/7 MPa) within three days to permit handling. Because fission product heating will be high, about 0.6 W/kg for a mortar containing 10% waste salt, the effects of elevated temperatures during curing and storage on mortar properties must be considered

  12. Alternatives for definse waste-salt disposal

    International Nuclear Information System (INIS)

    Benjamin, R.W.; McDonell, W.R.

    1983-01-01

    Alternatives for disposal of decontaminated high-level waste salt at Savannah River were reviewed to estimate costs and potential environmental impact for several processes. In this review, the reference process utilizing intermediate-depth burial of salt-concrete (saltcrete) monoliths was compared with alternatives including land application of the decontaminated salt as fertilizer for SRP pine stands, ocean disposal with and without containment, and terminal storage as saltcake in existing SRP waste tanks. Discounted total costs for the reference process and its modifications were in the same range as those for most of the alternative processes; uncontained ocean disposal with truck transport to Savannah River barges and storage as saltcake in SRP tanks had lower costs, but presented other difficulties. Environmental impacts could generally be maintained within acceptable limits for all processes except retention of saltcake in waste tanks, which could result in chemical contamination of surrounding areas on tank collapse. Land application would require additional salt decontamination to meet radioactive waste disposal standards, and ocean disposal without containment is not permitted in existing US practice. The reference process was judged to be the only salt disposal option studied which would meet all current requirements at an acceptable cost

  13. Waste salt recovery, recycle, and destruction

    International Nuclear Information System (INIS)

    Hickman, R.G.

    1992-12-01

    Starting in 1943 and continuing into the 1970s, radioactive wastes resulting from plutonium processing at Hanford were stored underground in 149 single shell tanks. Of these tanks, 66 are known or believedto be leaking, and over a period are believed to have leaked about 750,000 gal into the surrounding soil. The bulk of the aqueous solution has been removed and transferred to double shell tanks, none of which are leaking. The waste consists of 37 million gallons of salt cake and sludge. Most of the salt cake is sodium nitrate and other sodium salts. A substantial fraction of the sludge is sodium nitrate. Small amounts of the radionuclides are present in the sludge as oxides or hydroxides. In addition, some of the tanks contain organic compounds and ferrocyanide complexes, many of which have undergone radiolytic induced chemical changes during the years of storage. As part of the Hanford site remediation effort, the tank wastes must be removed, treated, and the residuals must be immobilized and disposed of in an environmentally acceptable manner. Removal methods of the waste from the tanks fall generally into three approaches: dry removal, slurry removal, and solution removed. The latter two methods are likely to result in some additional leakage to the surrounding soil, but that may be acceptable if the tank can be emptied and remediated before the leaked material permeates deeply into the soil. This effort includes three parts: salt splitting, acid separation, and destruction, with initial emphasis on salt splitting

  14. Waste isolation facility description: bedded salt

    Energy Technology Data Exchange (ETDEWEB)

    1976-09-01

    The waste isolation facility is designed to receive and store three basic types of solidified wastes: high-level wastes, intermediate level high-gamma transuranic waste, and low-gamma transuranic wastes. The facility under consideration in this report is designed for bedded salt at a depth of approximately 1800 ft. The present design for the facility includes an area which would be used initially as a pilot facility to test the viability of the concept, and a larger facility which would constitute the final storage area. The total storage area in the pilot facility is planned to be 77 acres and in the fuel facility 1601 acres. Other areas for shaft operations and access would raise the overall size of the total facility to slightly less than 2,000 acres. The following subjects are discussed in detail: surface facilities, shaft design and characteristics, design and construction of the underground waste isolation facility, ventilation systems, and design requirements and criteria. (LK)

  15. Waste isolation facility description: bedded salt

    International Nuclear Information System (INIS)

    1976-09-01

    The waste isolation facility is designed to receive and store three basic types of solidified wastes: high-level wastes, intermediate level high-gamma transuranic waste, and low-gamma transuranic wastes. The facility under consideration in this report is designed for bedded salt at a depth of approximately 1800 ft. The present design for the facility includes an area which would be used initially as a pilot facility to test the viability of the concept, and a larger facility which would constitute the final storage area. The total storage area in the pilot facility is planned to be 77 acres and in the fuel facility 1601 acres. Other areas for shaft operations and access would raise the overall size of the total facility to slightly less than 2,000 acres. The following subjects are discussed in detail: surface facilities, shaft design and characteristics, design and construction of the underground waste isolation facility, ventilation systems, and design requirements and criteria

  16. Immobilization of IFR salt wastes in mortar

    International Nuclear Information System (INIS)

    Fisher, D.F.; Johnson, T.R.

    1988-01-01

    Portland cement-base mortars are being considered for immobilizing chloride salt wastes from the fuel cycle of an integral fast reactor (IFR). The IFR is a sodium-cooled fast reactor with metal fuel. It has a close-coupled fuel cycle in which fission products are separated from the actinides in an electrochemical cell operating at 500 degrees C. This cell has a cadmium anode and a liquid salt electrolyte. The salt will be a low-melting mixture of alkaline and alkaline earth chlorides. This paper discusses one method being considered for immobilizing this treated salt, to disperse it in a portland cement-base motar, which would then be sealed in corrosion-resistant containers. For this application, the grout must be sufficiently fluid that it can be pumped into canisters where it will solidify into a strong, leach-resistant material

  17. Molten salt destruction process for mixed wastes

    International Nuclear Information System (INIS)

    Upadhye, R.S.; Wilder, J.G.; Karlsen, C.E.

    1993-04-01

    We are developing an advanced two-stage process for the treatment of mixed wastes, which contain both hazardous and radioactive components. The wastes, together with an oxidant gas, such as air, are injected into a bed of molten salt comprising a mixture of sodium-, potassium-, and lithium-carbonates, with a melting point of about 580 degree C. The organic constituents of the mixed waste are destroyed through the combined effect of pyrolysis and oxidation. Heteroatoms. such as chlorine, in the mixed waste form stable salts, such as sodium chloride, and are retained in the melt. The radioactive actinides in the mixed waste are also retained in the melt because of the combined action of wetting and partial dissolution. The original process, consists of a one-stage unit, operated at 900--1000 degree C. The advanced two-stage process has two stages, one for pyrolysis and one for oxidation. The pyrolysis stage is designed to operate at 700 degree C. The oxidation stage can be operated at a higher temperature, if necessary

  18. Cementitious Stabilization of Mixed Wastes with High Salt Loadings

    International Nuclear Information System (INIS)

    Spence, R.D.; Burgess, M.W.; Fedorov, V.V.; Downing, D.J.

    1999-01-01

    Salt loadings approaching 50 wt % were tolerated in cementitious waste forms that still met leach and strength criteria, addressing a Technology Deficiency of low salt loadings previously identified by the Mixed Waste Focus Area. A statistical design quantified the effect of different stabilizing ingredients and salt loading on performance at lower loadings, allowing selection of the more effective ingredients for studying the higher salt loadings. In general, the final waste form needed to consist of 25 wt % of the dry stabilizing ingredients to meet the criteria used and 25 wt % water to form a workable paste, leaving 50 wt % for waste solids. The salt loading depends on the salt content of the waste solids but could be as high as 50 wt % if all the waste solids are salt

  19. Different Methods for Conditioning Chloride Salt Wastes

    International Nuclear Information System (INIS)

    De Angelis, G.; Fedeli, C.; Capone, M.; Marzo, G.A.; Mariani, M.; Da Ros, M.; Giacobbo, F.; Macerata, E.; Giola, M.

    2015-01-01

    Three different methods have been used to condition chloride salt wastes coming from pyro-processes. Two of them allow to synthesise sodalite, a naturally occurring mineral containing chlorine: the former, starting from Zeolite 4A, which transforms the zeolite into sodalite; the latter, which starts from kaolinite, giving sodalite as well. In addition, a new matrix, termed SAP (SiO 2 -Al 2 O 3 -P 2 O 5 ), has been synthesised. It is able to form different mineral phases which occlude fission metals. The products from the different processes have been fully characterised. In particular the chemical durability of the final waste forms has been determined using the standard product consistency test. According to the results obtained, SAP seems to be a promising matrix for the incorporation of chloride salt wastes from pyro-processes. Financial support from the Nuclear Fission Safety Programme of the European Union (projects ACSEPT, contract FP7-CP-2007- 211 267, and SACSESS, Collaborative Project 323282), as well as from Italian Ministry for Economic Development (Accordo di Programma: Piano Annuale di Realizzazione 2008-2009) is gratefully acknowledged. (authors)

  20. Waste form dissolution in bedded salt

    International Nuclear Information System (INIS)

    Kaufman, A.M.

    1980-01-01

    A model was devised for waste dissolution in bedded salt, a hydrologically tight medium. For a typical Spent UnReprocessed Fuel (SURF) emplacement, the dissolution rate wll be diffusion limited and will rise to a steady state value after t/sub eq/ approx. = 250 (1+(1-epsilon 0 ) K/sub D//epsilon 0 ) (years) epsilon 0 is the overpack porosity and K/sub d/ is the overpack sorption coefficient. The steady state dissolution rate itself is dominated by the solubility of UO 2 . Steady state rates between 5 x 10 -5 and .5 (g/year) are achievable by SURF emplacements in bedded salt without overpack, and rates between 5 x 10 -7 and 5 x 10 -3 (g/year) with an overpack having porosity of 10 -2

  1. Molten salt combustion of radioactive wastes

    International Nuclear Information System (INIS)

    Grantham, L.F.; McKenzie, D.E.; Richards, W.L.; Oldenkamp, R.D.

    1976-01-01

    The Atomics International Molten Salt Combustion Process reduces the weight and volume of combustible β-γ contaminated transuranic waste by utilizing air in a molten salt medium to combust organic materials, to trap particulates, and to react chemically with any acidic gases produced during combustion. Typically, incomplete combustion products such as hydrocarbons and carbon monoxide are below detection limits (i.e., 3 ) is directly related to the sodium chloride vapor pressure of the melt; >80% of the particulate is sodium chloride. Essentially all metal oxides (combustion ash) are retained in the melt, e.g., >99.9% of the plutonium, >99.6% of the europium, and >99.9% of the ruthenium are retained in the melt. Both bench-scale radioactive and pilot scale (50 kg/hr) nonradioactive combustion tests have been completed with essentially the same results. Design of three combustors for industrial applications are underway

  2. Geologic disposal of nuclear wastes: salt's lead is challenged

    International Nuclear Information System (INIS)

    Kerr, R.A.

    1979-01-01

    The types of radioactive waste disposal sites available are outlined. The use of salt deposits and their advantages are discussed. The reasons for the selection of the present site for the Waste Isolation Pilot Plant are presented. The possibilities of using salt domes along the Gulf Coast and not-salt rocks as nuclear waste repositories are also discussed. The sea bed characteristics are described and advantages of this type of site selection are presented

  3. Cerebral salt wasting following traumatic brain injury

    Directory of Open Access Journals (Sweden)

    Peter Taylor

    2017-04-01

    Full Text Available Hyponatraemia is the most commonly encountered electrolyte disturbance in neurological high dependency and intensive care units. Cerebral salt wasting (CSW is the most elusive and challenging of the causes of hyponatraemia, and it is vital to distinguish it from the more familiar syndrome of inappropriate antidiuretic hormone (SIADH. Managing CSW requires correction of the intravascular volume depletion and hyponatraemia, as well as mitigation of on-going substantial sodium losses. Herein we describe a challenging case of CSW requiring large doses of hypertonic saline and the subsequent substantial benefit with the addition of fludrocortisone.

  4. Waste Isolation Pilot Plant Salt Decontamination Testing

    Energy Technology Data Exchange (ETDEWEB)

    Rick Demmer; Stephen Reese

    2014-09-01

    On February 14, 2014, americium and plutonium contamination was released in the Waste Isolation Pilot Plant (WIPP) salt caverns. At the request of WIPP’s operations contractor, Idaho National Laboratory (INL) personnel developed several methods of decontaminating WIPP salt, using surrogate contaminants and also americium (241Am). The effectiveness of the methods is evaluated qualitatively, and to the extent possible, quantitatively. One of the requirements of this effort was delivering initial results and recommendations within a few weeks. That requirement, in combination with the limited scope of the project, made in-depth analysis impractical in some instances. Of the methods tested (dry brushing, vacuum cleaning, water washing, strippable coatings, and mechanical grinding), the most practical seems to be water washing. Effectiveness is very high, and it is very easy and rapid to deploy. The amount of wastewater produced (2 L/m2) would be substantial and may not be easy to manage, but the method is the clear winner from a usability perspective. Removable surface contamination levels (smear results) from the strippable coating and water washing coupons found no residual removable contamination. Thus, whatever is left is likely adhered to (or trapped within) the salt. The other option that shows promise is the use of a fixative barrier. Bartlett Nuclear, Inc.’s Polymeric Barrier System (PBS) proved the most durable of the coatings tested. The coatings were not tested for contaminant entrapment, only for coating integrity and durability.

  5. Molten salt oxidation of organic hazardous waste with high salt content.

    Science.gov (United States)

    Lin, Chengqian; Chi, Yong; Jin, Yuqi; Jiang, Xuguang; Buekens, Alfons; Zhang, Qi; Chen, Jian

    2018-02-01

    Organic hazardous waste often contains some salt, owing to the widespread use of alkali salts during industrial manufacturing processes. These salts cause complications during the treatment of this type of waste. Molten salt oxidation is a flameless, robust thermal process, with inherent capability of destroying the organic constituents of wastes, while retaining the inorganic ingredients in the molten salt. In the present study, molten salt oxidation is employed for treating a typical organic hazardous waste with a high content of alkali salts. The hazardous waste derives from the production of thiotriazinone. Molten salt oxidation experiments have been conducted using a lab-scale molten salt oxidation reactor, and the emissions of CO, NO, SO 2 , HCl and dioxins are studied. Impacts are investigated from the composition of the molten salts, the types of feeding tube, the temperature of molten carbonates and the air factor. Results show that the waste can be oxidised effectively in a molten salt bath. Temperature of molten carbonates plays the most important role. With the temperature rising from 600 °C to 750 °C, the oxidation efficiency increases from 91.1% to 98.3%. Compared with the temperature, air factor has but a minor effect, as well as the composition of the molten salts and the type of feeding tube. The molten carbonates retain chlorine with an efficiency higher than 99.9% and the emissions of dioxins are below 8 pg TEQ g -1 sample. The present study shows that molten salt oxidation is a promising alternative for the disposal of organic hazardous wastes containing a high salt content.

  6. Release rates from waste packages in a salt repository

    International Nuclear Information System (INIS)

    Chambre, P.L.; Hwang, Y.; Lee, W.W.L.; Pigford, T.H.

    1987-06-01

    In this report we present estimates of radionuclide release rates from waste packages into salt. This conservative and bounding analysis shows that release rates from waste packages in salt are well below the US Nuclear Regulatory Commission's performance objectives for the engineered barrier system. 2 refs., 2 figs

  7. Mixed Waste Salt Encapsulation Using Polysiloxane - Final Report

    International Nuclear Information System (INIS)

    Miller, C.M.; Loomis, G.G.; Prewett, S.W.

    1997-01-01

    A proof-of-concept experimental study was performed to investigate the use of Orbit Technologies polysiloxane grouting material for encapsulation of U.S. Department of Energy mixed waste salts leading to a final waste form for disposal. Evaporator pond salt residues and other salt-like material contaminated with both radioactive isotopes and hazardous components are ubiquitous in the DOE complex and may exceed 250,000,000 kg of material. Current treatment involves mixing low waste percentages (less than 10% by mass salt) with cement or costly thermal treatment followed by cementation to the ash residue. The proposed technology involves simple mixing of the granular salt material (with relatively high waste loadings-greater than 50%) in a polysiloxane-based system that polymerizes to form a silicon-based polymer material. This study involved a mixing study to determine optimum waste loadings and compressive strengths of the resultant monoliths. Following the mixing study, durability testing was performed on promising waste forms. Leaching studies including the accelerated leach test and the toxicity characteristic leaching procedure were also performed on a high nitrate salt waste form. In addition to this testing, the waste form was examined by scanning electron microscope. Preliminary cost estimates for applying this technology to the DOE complex mixed waste salt problem is also given

  8. Salt disposal of heat-generating nuclear waste

    International Nuclear Information System (INIS)

    Leigh, Christi D.; Hansen, Francis D.

    2011-01-01

    This report summarizes the state of salt repository science, reviews many of the technical issues pertaining to disposal of heat-generating nuclear waste in salt, and proposes several avenues for future science-based activities to further the technical basis for disposal in salt. There are extensive salt formations in the forty-eight contiguous states, and many of them may be worthy of consideration for nuclear waste disposal. The United States has extensive experience in salt repository sciences, including an operating facility for disposal of transuranic wastes. The scientific background for salt disposal including laboratory and field tests at ambient and elevated temperature, principles of salt behavior, potential for fracture damage and its mitigation, seal systems, chemical conditions, advanced modeling capabilities and near-future developments, performance assessment processes, and international collaboration are all discussed. The discussion of salt disposal issues is brought current, including a summary of recent international workshops dedicated to high-level waste disposal in salt. Lessons learned from Sandia National Laboratories' experience on the Waste Isolation Pilot Plant and the Yucca Mountain Project as well as related salt experience with the Strategic Petroleum Reserve are applied in this assessment. Disposal of heat-generating nuclear waste in a suitable salt formation is attractive because the material is essentially impermeable, self-sealing, and thermally conductive. Conditions are chemically beneficial, and a significant experience base exists in understanding this environment. Within the period of institutional control, overburden pressure will seal fractures and provide a repository setting that limits radionuclide movement. A salt repository could potentially achieve total containment, with no releases to the environment in undisturbed scenarios for as long as the region is geologically stable. Much of the experience gained from United

  9. Salt disposal of heat-generating nuclear waste.

    Energy Technology Data Exchange (ETDEWEB)

    Leigh, Christi D. (Sandia National Laboratories, Carlsbad, NM); Hansen, Francis D.

    2011-01-01

    This report summarizes the state of salt repository science, reviews many of the technical issues pertaining to disposal of heat-generating nuclear waste in salt, and proposes several avenues for future science-based activities to further the technical basis for disposal in salt. There are extensive salt formations in the forty-eight contiguous states, and many of them may be worthy of consideration for nuclear waste disposal. The United States has extensive experience in salt repository sciences, including an operating facility for disposal of transuranic wastes. The scientific background for salt disposal including laboratory and field tests at ambient and elevated temperature, principles of salt behavior, potential for fracture damage and its mitigation, seal systems, chemical conditions, advanced modeling capabilities and near-future developments, performance assessment processes, and international collaboration are all discussed. The discussion of salt disposal issues is brought current, including a summary of recent international workshops dedicated to high-level waste disposal in salt. Lessons learned from Sandia National Laboratories' experience on the Waste Isolation Pilot Plant and the Yucca Mountain Project as well as related salt experience with the Strategic Petroleum Reserve are applied in this assessment. Disposal of heat-generating nuclear waste in a suitable salt formation is attractive because the material is essentially impermeable, self-sealing, and thermally conductive. Conditions are chemically beneficial, and a significant experience base exists in understanding this environment. Within the period of institutional control, overburden pressure will seal fractures and provide a repository setting that limits radionuclide movement. A salt repository could potentially achieve total containment, with no releases to the environment in undisturbed scenarios for as long as the region is geologically stable. Much of the experience gained from

  10. Radioactive waste and special waste disposal in salt domes - phoney waste management solutions

    International Nuclear Information System (INIS)

    Grimmel, E.

    1990-01-01

    The paper tries to make aware of the fact that an indefinite safe disposal of anthropogeneous wastes in underground repositories is impossible. Suspicion is raised that the Gorleben-Rambow salt dome has never been studied for its suitability as a repository, but that it was simply taken for granted. Safety analyses are meant only to conceal uncertainty. It is demanded to immediately opt out of the ultimate disposal technique for radioactive and special wastes in salt caverns. (DG) [de

  11. Projected Salt Waste Production from a Commercial Pyroprocessing Facility

    Directory of Open Access Journals (Sweden)

    Michael F. Simpson

    2013-01-01

    Full Text Available Pyroprocessing of used nuclear fuel inevitably produces salt waste from electrorefining and/or oxide reduction unit operations. Various process design characteristics can affect the actual mass of such waste produced. This paper examines both oxide and metal fuel treatment, estimates the amount of salt waste generated, and assesses potential benefit of process options to mitigate the generation of salt waste. For reference purposes, a facility is considered in which 100 MT/year of fuel is processed. Salt waste estimates range from 8 to 20 MT/year from considering numerous scenarios. It appears that some benefit may be derived from advanced processes for separating fission products from molten salt waste, but the degree of improvement is limited. Waste form production is also considered but appears to be economically unfavorable. Direct disposal of salt into a salt basin type repository is found to be the most promising with respect to minimizing the impact of waste generation on the economic feasibility and sustainability of pyroprocessing.

  12. Effect of indomethacin and salt depletion on renal proton MR imaging

    International Nuclear Information System (INIS)

    Heyman, S.N.; Mammen, M.

    1991-01-01

    Blockade of the synthesis of vasodilating prostaglandins with non-steroidal anti-inflammatory drugs (NSAID) renders the renal medulla susceptible to hypoxic injury with reduced renal function, especially in clinical conditions characterized by volume depletion. Alterations in renal hemodynamics and urine production may effect renal MR imaging under these circumstances. We injected salt-depleted and control rats undergoing proton MR imaging with indomethacin 10 mg/kg. Indomethacin abolished the cortico-medullary T2-gradient and markedly diminished the overall renal signal in salt-depleted rats only. These changes, which progressed over a period of 40 min after indomethacin was injected, probably result from renal oligemia and decreased urine production, with an associated decrease in T2-values. We suggest that a history of consumption of non-steroidal anti-inflammatory drugs should be obtained and taken into account in the evaluation of renal proton MR imaging, especially in the presence of salt and volume depletion. (orig.)

  13. Defense waste salt disposal at the Savannah River Plant

    International Nuclear Information System (INIS)

    Langton, C.A.; Dukes, M.D.

    1984-01-01

    A cement-based waste form, saltstone, has been designed for disposal of Savannah River Plant low-level radioactive salt waste. The disposal process includes emplacing the saltstone in engineered trenches above the water table but below grade at SRP. Design of the waste form and disposal system limits the concentration of salts and radionuclides in the groundwater so that EPA drinking water standards will not be exceeded at the perimeter of the disposal site. 10 references, 4 figures, 3 tables

  14. Expected brine movement at potential nuclear waste repository salt sites

    International Nuclear Information System (INIS)

    McCauley, V.S.; Raines, G.E.

    1987-08-01

    The BRINEMIG brine migration code predicts rates and quantities of brine migration to a waste package emplaced in a high-level nuclear waste repository in salt. The BRINEMIG code is an explicit time-marching finite-difference code that solves a mass balance equation and uses the Jenks equation to predict velocities of brine migration. Predictions were made for the seven potentially acceptable salt sites under consideration as locations for the first US high-level nuclear waste repository. Predicted total quantities of accumulated brine were on the order of 1 m 3 brine per waste package or less. Less brine accumulation is expected at domal salt sites because of the lower initial moisture contents relative to bedded salt sites. Less total accumulation of brine is predicted for spent fuel than for commercial high-level waste because of the lower temperatures generated by spent fuel. 11 refs., 36 figs., 29 tabs

  15. Alternative methods of salt disposal at the seven salt sites for a nuclear waste repository

    International Nuclear Information System (INIS)

    1987-02-01

    This study discusses the various alternative salt management techniques for the disposal of excess mined salt at seven potentially acceptable nuclear waste repository sites: Deaf Smith and Swisher Counties, Texas; Richton and Cypress Creek Domes, Mississippi; Vacherie Dome, Louisiana; and Davis and Lavender Canyons, Utah. Because the repository development involves the underground excavation of corridors and waste emplacement rooms, in either bedded or domed salt formations, excess salt will be mined and must be disposed of offsite. The salt disposal alternatives examined for all the sites include commercial use, ocean disposal, deep well injection, landfill disposal, and underground mine disposal. These alternatives (and other site-specific disposal methods) are reviewed, using estimated amounts of excavated, backfilled, and excess salt. Methods of transporting the excess salt are discussed, along with possible impacts of each disposal method and potential regulatory requirements. A preferred method of disposal is recommended for each potentially acceptable repository site. 14 refs., 5 tabs

  16. High-temperature vacuum distillation separation of plutonium waste salts

    International Nuclear Information System (INIS)

    Garcia, E.

    1996-01-01

    In this task, high-temperature vacuum distillation separation is being developed for residue sodium chloride-potassium chloride salts resulting from past pyrochemical processing of plutonium. This process has the potential of providing clean separation of the salt and the actinides with minimal amounts of secondary waste generation. The process could produce chloride salt that could be discarded as low-level waste (LLW) or low actinide content transuranic (TRU) waste, and a concentrated actinide oxide powder that would meet long-term storage standards (DOE-DTD-3013-94) until a final disposition option for all surplus plutonium is chosen

  17. Expected environment for waste packages in a salt repository

    International Nuclear Information System (INIS)

    Pederson, L.R.; Clark, D.E.; Hodges, F.N.; McVay, G.L.; Rai, D.

    1983-01-01

    This paper discusses results of recent efforts to define the very near-field (within approximately 2 m) environmental conditions to which waste packages will be exposed in a salt repository. These conditions must be considered in the experimental design for waste package materials testing, which includes corrosion of barrier materials and leaching of waste forms. Site-specific brine compositions have been determined, and standard brine compositions have been selected for testing purposes. Actual brine compositions will vary depending on origin, temperature, irradiation history, and contact with irradiated rock salt. Results of irradiating rock salt, synthetic brines, rock salt/brine mixtures, and reactions of irradiated rock salt with brine solutions are reported. 38 references, 3 figures, 2 tables

  18. Roadmap for disposal of Electrorefiner Salt as Transuranic Waste.

    Energy Technology Data Exchange (ETDEWEB)

    Rechard, Robert P. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Trone, Janis R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Kalinina, Elena Arkadievna [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wang, Yifeng [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Hadgu, Teklu [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sanchez, Lawrence C. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-12-01

    The experimental breeder reactor (EBR-II) used fuel with a layer of sodium surrounding the uranium-zirconium fuel to improve heat transfer. Disposing of EBR-II fuel in a geologic repository without treatment is not prudent because of the potentially energetic reaction of the sodium with water. In 2000, the US Department of Energy (DOE) decided to treat the sodium-bonded fuel with an electrorefiner (ER), which produces metallic uranium product, a metallic waste, mostly from the cladding, and the salt waste in the ER, which contains most of the actinides and fission products. Two waste forms were proposed for disposal in a mined repository; the metallic waste, which was to be cast into ingots, and the ER salt waste, which was to be further treated to produce a ceramic waste form. However, alternative disposal pathways for metallic and salt waste streams may reduce the complexity. For example, performance assessments show that geologic repositories can easily accommodate the ER salt waste without treating it to form a ceramic waste form. Because EBR-II was used for atomic energy defense activities, the treated waste likely meets the definition of transuranic waste. Hence, disposal at the Waste Isolation Pilot Plant (WIPP) in southern New Mexico, may be feasible. This report reviews the direct disposal pathway for ER salt waste and describes eleven tasks necessary for implementing disposal at WIPP, provided space is available, DOE decides to use this alternative disposal pathway in an updated environmental impact statement, and the State of New Mexico grants permission.

  19. Potential for creation of a salt dome following disposal of radioactive waste in a salt layer

    International Nuclear Information System (INIS)

    Charo, L.; Habib, P.

    1987-01-01

    The study aims at quantifying the possibility of creation of a salt dome from a salt layer in which heat-emitting radioactive waste would be buried. Volume 1 describes the results of numerical computer simulations, and of laboratory-scale models in centrifuges. Volume 2 envisages, in a geological perspective, the origin of salt domes, the mechanisms of their formation, and the associated parameters [fr

  20. Potential for creation of a salt dome following disposal of radioactive waste in a salt layer

    International Nuclear Information System (INIS)

    Fries, G.

    1987-01-01

    The study aims at quantifying the possibility of creation of a salt dome from a salt layer in which heat-emitting radioactive waste would be buried. Volume 1 describes the results of numerical computer simulations, and of laboratory-scale models in centrifuges. Volume 2 envisages, in a geological perspective, the origin of salt domes, the mechanisms of thei formation, and the associated parameters [fr

  1. High-salt diets during pregnancy affected fetal and offspring renal renin-angiotensin system.

    Science.gov (United States)

    Mao, Caiping; Liu, Rong; Bo, Le; Chen, Ningjing; Li, Shigang; Xia, Shuixiu; Chen, Jie; Li, Dawei; Zhang, Lubo; Xu, Zhice

    2013-07-01

    Intrauterine environments are related to fetal renal development and postnatal health. Influence of salty diets during pregnancy on renal functions and renin-angiotensin system (RAS) was determined in the ovine fetuses and offspring. Pregnant ewes were fed high-salt diet (HSD) or normal-salt diet (NSD) for 2 months during middle-to-late gestation. Fetal renal functions, plasma hormones, and mRNA and protein expressions of the key elements of renal RAS were measured in the fetuses and offspring. Fetal renal excretion of sodium was increased while urine volume decreased in the HSD group. Fetal blood urea nitrogen was increased, while kidney weight:body weight ratio decreased in the HSD group. The altered ratio was also observed in the offspring aged 15 and 90 days. Maternal and fetal plasma antidiuretic hormone was elevated without changes in plasma renin activity and Ang I levels, while plasma Ang II was decreased. The key elements of local renal RAS, including angiotensinogen, angiotensin converting enzyme (ACE), ACE2, AT1, and AT2 receptor expression in both mRNA and protein, except renin, were altered following maternal high salt intake. The results suggest that high intake of salt during pregnancy affected fetal renal development associated with an altered expression of the renal key elements of RAS, some alterations of fetal origins remained after birth as possible risks in developing renal or cardiovascular diseases.

  2. Treatment of waste salts by oxygen sparging and vacuum distillation

    International Nuclear Information System (INIS)

    Cho, Y.J.; Yang, H.C.; Kim, E.H.; Kin, I.T.; Eun, H.C.

    2007-01-01

    Full text of publication follows. During the electrorefining process of the oxide spent fuel from LWR, amounts of waste salts containing some metal chloride species such as rare earths and actinide chlorides are generated, where the reuse of the waste salts is very important from the standpoint of an economical as well as an environmental aspect. In order to reuse the waste salts, a salt vacuum distillation method can be used. For the best separation by a vacuum distillation, the metal chloride species involved in the waste salts must be converted into their oxide(or oxychloride) forms due to the their low volatility compared to that of LiCl-KCl. In this study, an oxygen sparging process was adopted for the oxidation (or precipitation) of rare earth chlorides. The effects of oxygen flow rate and molten salt temperature on the conversion of rare earth chlorides to the precipitate phase (i.e. oxide or oxychloride) were investigated. In addition, distillation characteristics of LiCl-KCl molten salt with system pressure and temperature were studied. (authors)

  3. Salt removal from tanks containing high-level radioactive waste

    International Nuclear Information System (INIS)

    Kiser, D.L.

    1981-01-01

    At the Savannah River Plant (SRP), there are 23 waste storage tanks containing high-level radioactive wastes that are to be retired. These tanks contain about 23 million liters of salt and about 10 million liters of sludge, that are to be relocated to new Type III, fully stress-relieved tanks with complete secondary containment. About 19 million liters of salt cake are to be dissolved. Steam jet circulators were originally proposed for the salt dissolution program. However, use of steam jet circulators raised the temperature of the tank contents and caused operating problems. These included increased corrosion risk and required long cooldown periods prior to transfer. Alternative dissolution concepts were investigated. Examination of mechanisms affecting salt dissolution showed that the ability of fresh water to contact the cake surface was the most significant factor influencing dissolution rate. Density driven and mechanical agitation techniques were developed on a bench scale and then were demonstrated in an actual waste tank. Actual waste tank demonstrations were in good agreement with bench-scale experiments at 1/85 scale. The density driven method utilizes simple equipment, but leaves a cake heel in the tank and is hindered by the presence of sludge or Zeolite in the salt cake. Mechanical agitation overcomes the problems found with both steam jet circulators and the density driven technique and is the best method for future waste tank salt removal

  4. Waste salt disposal at the Savannah River Plant

    International Nuclear Information System (INIS)

    Langton, C.A.; Oblath, S.B.; Pepper, D.W.; Wilhite, E.L.

    1986-01-01

    Waste salt solution, produced during processing of high-level nuclear waste, will be incorporated in a cement matrix for emplacement in an engineered disposal facility. Wasteform characteristics and disposal facility details will be presented along with results of a field test of wasteform contaminant release and of modeling studies to predict releases. 5 refs., 11 figs., 5 tabs

  5. Waste package for a repository located in salt

    International Nuclear Information System (INIS)

    Basham, S.J. Jr.

    1983-01-01

    This paper describes the current status of the waste package designs for salt repositories. The status of the supporting studies of environment definition, corrosion of containment materials, and leaching of waste forms is also presented. Emphasis is on the results obtained in FY 83 and the planned effort in FY 84. 8 references, 3 figures, 1 table

  6. Moderate (20%) fructose-enriched diet stimulates salt-sensitive hypertension with increased salt retention and decreased renal nitric oxide.

    Science.gov (United States)

    Gordish, Kevin L; Kassem, Kamal M; Ortiz, Pablo A; Beierwaltes, William H

    2017-04-01

    Previously, we reported that 20% fructose diet causes salt-sensitive hypertension. In this study, we hypothesized that a high salt diet supplemented with 20% fructose (in drinking water) stimulates salt-sensitive hypertension by increasing salt retention through decreasing renal nitric oxide. Rats in metabolic cages consumed normal rat chow for 5 days (baseline), then either: (1) normal salt for 2 weeks, (2) 20% fructose in drinking water for 2 weeks, (3) 20% fructose for 1 week, then fructose + high salt (4% NaCl) for 1 week, (4) normal chow for 1 week, then high salt for 1 week, (5) 20% glucose for 1 week, then glucose + high salt for 1 week. Blood pressure, sodium excretion, and cumulative sodium balance were measured. Systolic blood pressure was unchanged by 20% fructose or high salt diet. 20% fructose + high salt increased systolic blood pressure from 125 ± 1 to 140 ± 2 mmHg ( P  fructose + high salt than either high salt, or glucose + high salt (114.2 ± 4.4 vs. 103.6 ± 2.2 and 98.6 ± 5.6 mEq/Day19; P  fructose + high salt group compared to high salt only: 5.33 ± 0.21 versus 7.67 ± 0.31 mmol/24 h; P  fructose + high salt group (2139 ± 178  μ mol /24 hrs P  fructose predisposes rats to salt-sensitivity and, combined with a high salt diet, leads to sodium retention, increased blood pressure, and impaired renal nitric oxide availability. © 2017 The Authors. Physiological Reports published by Wiley Periodicals, Inc. on behalf of The Physiological Society and the American Physiological Society.

  7. Molten salt hazardous waste disposal process utilizing gas/liquid contact for salt recovery

    International Nuclear Information System (INIS)

    Grantham, L.F.; McKenzie, D.E.

    1984-01-01

    The products of a molten salt combustion of hazardous wastes are converted into a cooled gas, which can be filtered to remove hazardous particulate material, and a dry flowable mixture of salts, which can be recycled for use in the molten salt combustion, by means of gas/liquid contact between the gaseous products of combustion of the hazardous waste and a solution produced by quenching the spent melt from such molten salt combustion. The process results in maximizing the proportion of useful materials recovered from the molten salt combustion and minimizing the volume of material which must be discarded. In a preferred embodiment a spray dryer treatment is used to achieve the desired gas/liquid contact

  8. Molten salt treatment to minimize and optimize waste

    International Nuclear Information System (INIS)

    Gat, U.; Crosley, S.M.; Gay, R.L.

    1993-01-01

    A combination molten salt oxidizer (MSO) and molten salt reactor (MSR) is described for treatment of waste. The MSO is proposed for contained oxidization of organic hazardous waste, for reduction of mass and volume of dilute waste by evaporation of the water. The NTSO residue is to be treated to optimize the waste in terms of its composition, chemical form, mixture, concentration, encapsulation, shape, size, and configuration. Accumulations and storage are minimized, shipments are sized for low risk. Actinides, fissile material, and long-lived isotopes are separated and completely burned or transmuted in an MSR. The MSR requires no fuel element fabrication, accepts the materials as salts in arbitrarily small quantities enhancing safety, security, and overall acceptability

  9. Assessment of crushed salt consolidation and fracture healing processes in a nuclear waste repository in salt

    International Nuclear Information System (INIS)

    1984-11-01

    For a nuclear waste repository in salt, two aspects of salt behavior are expected to contribute to favorable conditions for waste isolation. First, consolidation of crushed salt backfill due to creep closure of the underground openings may result in a backfill barrier with low permeability. Second, fractures created in the salt by excavation may heal under the influence of stress and temperature following sealing. This report reviews the status of knowledge regarding crushed salt consolidation and fracture healing, provides analyses which predict the rates at which the processes will occur under repository conditions, and develops requirements for future study. Analyses of the rate at which crushed salt will consolidate are found to be uncertain because of unexplained wide variation in the creep properties of crushed salt obtained from laboratory testing, and because of uncertainties in predictions of long term closure rates of openings in salt. This uncertainty could be resolved to a large degree by additional laboratory testing of crushed salt. Similarly, additional testing of fracture healing processes is required to confirm that healing will be effective under repository conditions. Extensive references, 27 figures, 5 tables

  10. Mineralocorticoid-induced sodium appetite and renal salt retention: Evidence for common signaling and effector mechanisms

    Science.gov (United States)

    Fu, Yiling; Vallon, Volker

    2014-01-01

    An increase in renal sodium chloride (salt) retention and an increase in sodium appetite is the body's response to salt restriction or depletion in order to restore salt balance. Renal salt retention and increased sodium appetite can also be maladaptive and sustain the pathophysiology in conditions like salt-sensitive hypertension and chronic heart failure. Here we review the central role of the mineralocorticoid aldosterone in both the increase in renal salt reabsorption and sodium appetite. We discuss the working hypothesis that aldosterone activates similar signaling and effector mechanisms in the kidney and brain, including the mineralocorticoid receptor, the serum-and-glucocorticoid-induced kinase SGK1, the ubiquitin ligase NEDD4-2, and the epithelial sodium channel ENaC. The latter also mediates the gustatory salt sensing in the tongue, which is required for the manifestation of increased salt intake. Effects of aldosterone on both brain and kidney synergize with the effects of angiotensin II. Thus, mineralocorticoids appear to induce similar molecular pathways in the kidney, brain, and possibly tongue, which could provide opportunities for more effective therapeutic interventions. Inhibition of renal salt reabsorption is compensated by stimulation of salt appetite and vice versa; targeting both mechanisms should be more effective. Inhibiting the arousal to consume salty food may improve a patient's compliance to reducing salt intake. While a better understanding of the molecular mechanisms is needed and will provide new options, current pharmacological interventions that target both salt retention and sodium appetite include mineralocorticoid receptor antagonists and potentially inhibitors of angiotensin II and ENaC. PMID:25376899

  11. The safe disposal of radioactive wastes in geologic salt formations

    International Nuclear Information System (INIS)

    Kuehn, K.; Proske, R.

    Geologic salt formations appear to be particularly suitable for final storage. Their existance alone - the salt formations in Northern Germany are more than 200 million years old - is proof of their stability and of their isolation from biological cycles. In 1967 the storage of LAW and later, in 1972, of MAW was started in the experimental storage area Asse, south-east of Braunschweig, after the necessary technical preparations had been made. In more than ten years of operation approx. 114,000 drums of slightly active and 1,298 drums of medium-active wastes were deposited without incident. Methods have been developed for filling the available caverns with wastes and salt to ensure the security of long term disposal without supervision. Tests with electric heaters for simulation of heat-generating highly active wastes confirm the good suitability of salt formations for storing these wastes. Safety analyses for the operating time as well as for the long term phase after closure of the final storage area, which among others also comprise the improbable ''greatest expected accident'', namely break through of water, are carried out and confirm the safety of ultimate storage of radioactive wastes in geological salt formations. (orig./HP) [de

  12. Salt Repository Project Waste Package Program Plan: Draft

    International Nuclear Information System (INIS)

    Carr, J.A.; Cunnane, J.C.

    1986-01-01

    Under the direction of the Office of Civilian Radioactive Waste Management (OCRWM) created within the DOE by direction of the Nuclear Waste Policy Act of 1982 (NWPA), the mission of the Salt Repository Project (SRP) is to provide for the development of a candidate salt repository for disposal of high-level radioactive waste (HLW) and spent reactor fuel in a manner that fully protects the health and safety of the public and the quality of the environment. In consideration of the program needs and requirements discussed above, the SRP has decided to develop and issue this SRP Waste Package Program Plan. This document is intended to outline how the SRP plans to develop the waste package design and to show, with reasonable assurance, that the developed design will satisfy applicable requirements/performance objectives. 44 refs., 16 figs., 16 tabs

  13. Membrane Treatment of Liquid Salt Bearing Radioactive Wastes

    International Nuclear Information System (INIS)

    Dmitriev, S. A.; Adamovich, D. V.; Demkin, V. I.; Timofeev, E. M.

    2003-01-01

    The main fields of introduction and application of membrane methods for preliminary treatment and processing salt liquid radioactive waste (SLRW) can be nuclear power stations (NPP) and enterprises on atomic submarines (AS) utilization. Unlike the earlier developed technology for the liquid salt bearing radioactive waste decontamination and concentrating this report presents the new enhanced membrane technology for the liquid salt bearing radioactive waste processing based on the state-of-the-art membrane unit design, namely, the filtering units equipped with the metal-ceramic membranes of ''TruMem'' brand, as well as the electrodialysis and electroosmosis concentrators. Application of the above mentioned units in conjunction with the pulse pole changer will allow the marked increase of the radioactive waste concentrating factor and the significant reduction of the waste volume intended for conversion into monolith and disposal. Besides, the application of the electrodialysis units loaded with an ion exchange material at the end polishing stage of the radioactive waste decontamination process will allow the reagent-free radioactive waste treatment that meets the standards set for the release of the decontaminated liquid radioactive waste effluents into the natural reservoirs of fish-farming value

  14. Molten salt processing of mixed wastes with offgas condensation

    International Nuclear Information System (INIS)

    Cooper, J.F.; Brummond, W.; Celeste, J.; Farmer, J.; Hoenig, C.; Krikorian, O.H.; Upadhye, R.; Gay, R.L.; Stewart, A.; Yosim, S.

    1991-01-01

    We are developing an advanced process for treatment of mixed wastes in molten salt media at temperatures of 700--1000 degrees C. Waste destruction has been demonstrated in a single stage oxidation process, with destruction efficiencies above 99.9999% for many waste categories. The molten salt provides a heat transfer medium, prevents thermal surges, and functions as an in situ scrubber to transform the acid-gas forming components of the waste into neutral salts and immobilizes potentially fugitive materials by a combination of particle wetting, encapsulation and chemical dissolution and solvation. Because the offgas is collected and assayed before release, and wastes containing toxic and radioactive materials are treated while immobilized in a condensed phase, the process avoids the problems sometimes associated with incineration processes. We are studying a potentially improved modification of this process, which treats oxidizable wastes in two stages: pyrolysis followed by catalyzed molten salt oxidation of the pyrolysis gases at ca. 700 degrees C. 15 refs., 5 figs., 1 tab

  15. Integrated demonstration of molten salt oxidation with salt recycle for mixed waste treatment

    International Nuclear Information System (INIS)

    Hsu, P.C.

    1997-01-01

    Molten Salt Oxidation (MSO) is a thermal, nonflame process that has the inherent capability of completely destroying organic constituents of mixed wastes, hazardous wastes, and energetic materials while retaining inorganic and radioactive constituents in the salt. For this reason, MSO is considered a promising alternative to incineration for the treatment of a variety of organic wastes. Lawrence Livermore National Laboratory (LLNL) has prepared a facility and constructed an integrated pilot-scale MSO treatment system in which tests and demonstrations are performed under carefully controlled (experimental) conditions. The system consists of a MSO processor with dedicated off-gas treatment, a salt recycle system, feed preparation equipment, and equipment for preparing ceramic final waste forms. This integrated system was designed and engineered based on laboratory experience with a smaller engineering-scale reactor unit and extensive laboratory development on salt recycle and final forms preparation. In this paper we present design and engineering details of the system and discuss its capabilities as well as preliminary process demonstration data. A primary purpose of these demonstrations is identification of the most suitable waste streams and waste types for MSO treatment

  16. Thermal denitration of high concentration nitrate salts waste water

    International Nuclear Information System (INIS)

    Hwang, D. S.; Oh, J. H.; Choi, Y. D.; Hwang, S. T.; Park, J. H.; Latge, C.

    2003-01-01

    This study investigated the thermodynamic and the thermal decomposition properties of high concentration nitrate salts waste water for the lagoon sludge treatment. The thermodynamic property was carried out by COACH and GEMINI II based on the composition of nitrate salts waste water. The thermal decomposition property was carried out by TG-DTA and XRD. Ammonium nitrate and sodium nitrate were decomposed at 250 .deg. C and 730 . deg. C, respectively. Sodium nitrate could be decomposed at 450 .deg. C in the case of adding alumina for converting unstable Na 2 O into stable Na 2 O.Al 2 O 3 . The flow sheet for nitrate salts waste water treatment was proposed based on the these properties data. These will be used by the basic data of the process simulation

  17. Molt salts reactors capacity for wastes incineration and energy production

    International Nuclear Information System (INIS)

    David, S.; Nuttin, A.

    2005-01-01

    The molten salt reactors present many advantages in the framework of the IV generation systems development for the energy production and/or the wastes incineration. After a recall of the main studies realized on the molten salt reactors, this document presents the new concepts and the identified research axis: the MSRE project and experience, the incinerators concepts, the thorium cycle. (A.L.B.)

  18. Hydrological methods preferentially recover cesium from nuclear waste salt cake

    International Nuclear Information System (INIS)

    Brooke, J.N.; Hamm, L.L.

    1997-01-01

    The Savannah River Site is treating high level radioactive waste in the form of insoluble solids (sludge), crystallized salt (salt cake), and salt solutions. High costs and operational concerns have prompted DOE to look for ways to improve the salt cake treatment process. A numerical model was developed to evaluate the feasibility of pump and treat technology for extracting cesium from salt cake. A modified version of the VAM3DCG code was used to first establish a steady-state flow field, then to simulate 30 days of operation. Simulation results suggest that efficient cesium extraction can be obtained with low displacement volumes. The actual extraction process will probably be less impressive because of nonuniform properties. 2 refs., 2 figs

  19. Novel waste printed circuit board recycling process with molten salt

    OpenAIRE

    Riedewald, Frank; Sousa-Gallagher, Maria

    2015-01-01

    The objective of the method was to prove the concept of a novel waste PCBs recycling process which uses inert, stable molten salts as the direct heat transfer fluid and, simultaneously, uses this molten salt to separate the metal products in either liquid (solder, zinc, tin, lead, etc.) or solid (copper, gold, steel, palladium, etc.) form at the operating temperatures of 450?470??C. The PCB recovery reactor is essentially a U-shaped reactor with the molten salt providing a continuous fluid, a...

  20. Containment of solidified liquid hazardous waste in domal salt

    International Nuclear Information System (INIS)

    Domenico, P.A.; Lerman, A.

    1992-01-01

    In recent years, the solidification of hazardous liquid waste has become a viable option in waste management. The solidification process results in an increased volume but more stable waste form that must be disposed of or stored in a dry environment. An environment of choice in south central Texas is domal salt. The salt dome currently under investigation has a water content of 0.002 percent by weight and a permeability less than one nanodarcy. A question that must be addressed is whether a salt dome has a particular set of attributes that will prevent the release of contaminants to the environment. From a regulatory perspective, a ''no migration'' petition must be approved by the U.S.E.P.A. for the containment facility. By ''no migration'' it is implied that the waste must be contained for 10,000 years. A demonstration that this condition will be met will require model calculations and such models must be based on the physical and chemical characteristics of the waste form and the geologic environment. In particular, the models must address the rate of brine infiltration into the caverns, providing information on how fast an immobile solid waste form could convert to a more mobile liquid state. Additionally, the potential for migration by both diffusion and advection is of concern. Lastly, given a partially saturated cavern, the question of how far gaseous waste will be transported over the 10,000 year containment period must also be addressed. Results indicate that the containment capabilities of domal salt are exceptional. A nominal volume of brine will seep into the cavern and most voids between the injected solidified waste pellets will remain unsaturated. Very small quantities of hazardous constituents will be leached from the waste pellets

  1. Mixing Modeling Analysis For SRS Salt Waste Disposition

    International Nuclear Information System (INIS)

    Lee, S.

    2011-01-01

    Nuclear waste at Savannah River Site (SRS) waste tanks consists of three different types of waste forms. They are the lighter salt solutions referred to as supernate, the precipitated salts as salt cake, and heavier fine solids as sludge. The sludge is settled on the tank floor. About half of the residual waste radioactivity is contained in the sludge, which is only about 8 percentage of the total waste volume. Mixing study to be evaluated here for the Salt Disposition Integration (SDI) project focuses on supernate preparations in waste tanks prior to transfer to the Salt Waste Processing Facility (SWPF) feed tank. The methods to mix and blend the contents of the SRS blend tanks were evalutaed to ensure that the contents are properly blended before they are transferred from the blend tank such as Tank 50H to the SWPF feed tank. The work consists of two principal objectives to investigate two different pumps. One objective is to identify a suitable pumping arrangement that will adequately blend/mix two miscible liquids to obtain a uniform composition in the tank with a minimum level of sludge solid particulate in suspension. The other is to estimate the elevation in the tank at which the transfer pump inlet should be located where the solid concentration of the entrained fluid remains below the acceptance criterion (0.09 wt% or 1200 mg/liter) during transfer operation to the SWPF. Tank 50H is a Waste Tank that will be used to prepare batches of salt feed for SWPF. The salt feed must be a homogeneous solution satisfying the acceptance criterion of the solids entrainment during transfer operation. The work described here consists of two modeling areas. They are the mixing modeling analysis during miscible liquid blending operation, and the flow pattern analysis during transfer operation of the blended liquid. The modeling results will provide quantitative design and operation information during the mixing/blending process and the transfer operation of the blended

  2. Laboratory simulation of salt dissolution during waste removal

    International Nuclear Information System (INIS)

    Wiersma, B.J.; Parish, W.R.

    1997-01-01

    Laboratory experiments were performed to support the field demonstration of improved techniques for salt dissolution in waste tanks at the Savannah River Site. The tests were designed to investigate three density driven techniques for salt dissolution: (1) Drain-Add-Sit-Remove, (2) Modified Density Gradient, and (3) Continuous Salt Mining. Salt dissolution was observed to be a very rapid process as salt solutions with densities between 1.38-1.4 were frequently removed. Slower addition and removal rates and locating the outlet line at deeper levels below the top of the saltcake provided the best contact between the dissolution water and the saltcake. It was observed that dissolution with 1 M sodium hydroxide solution resulted in salt solutions that were within the current inhibitor requirements for the prevention of stress corrosion cracking. This result was independent of the density driven technique. However, if inhibited water (0.01 M sodium hydroxide and 0.011 M sodium nitrite) was utilized, the salt solutions were frequently outside the inhibitor requirements. Corrosion testing at conditions similar to the environments expected during waste removal was recommended

  3. Waste management analysis for the nuclear fuel cycle. I. Actinide recovery from aqueous salt wastes

    International Nuclear Information System (INIS)

    Martella, L.L.; Navratil, J.D.

    1979-01-01

    A preliminary feasibility study of solvent extraction methods has been completed for removing actinides from selected salt wastes likely to be produced during reactor fuel fabrication and reprocessing. The use of a two-step solvent extraction system, tributyl phosphate (TBP) followed by a bidentate organophosphorus extractant (DHDECMP), appears most efficient for removing actinides from salt waste. The TBP step would remove most of the plutonium and >99.99% of the uranium. The second step, using DHDECMP, would remove >99.91% of the americium, the remaining plutonium (>99.98%), and other actinides from the acidified salt waste

  4. Characteristics of solidified products containing radioactive molten salt waste.

    Science.gov (United States)

    Park, Hwan-Seo; Kim, In-Tae; Cho, Yong-Zun; Eun, Hee-Chul; Kim, Joon-Hyung

    2007-11-01

    The molten salt waste from a pyroprocess to recover uranium and transuranic elements is one of the problematic radioactive wastes to be solidified into a durable wasteform for its final disposal. By using a novel method, named as the GRSS (gel-route stabilization/solidification) method, a molten salt waste was treated to produce a unique wasteform. A borosilicate glass as a chemical binder dissolves the silicate compounds in the gel products to produce one amorphous phase while most of the phosphates are encapsulated by the vitrified phase. Also, Cs in the gel product is preferentially situated in the silicate phase, and it is vitrified into a glassy phase after a heat treatment. The Sr-containing phase is mainly phosphate compounds and encapsulated by the glassy phase. These phenomena could be identified by the static and dynamic leaching test that revealed a high leach resistance of radionuclides. The leach rates were about 10(-3) - 10(-2) g/m2 x day for Cs and 10(-4) - 10(-3) g/m2 x day for Sr, and the leached fractions of them were predicted to be 0.89% and 0.39% at 900 days, respectively. This paper describes the characteristics of a unique wasteform containing a molten salt waste and provides important information on a newly developed immobilization technology for salt wastes, the GRSS method.

  5. Engineering Options Assessment Report. Nitrate Salt Waste Stream Processing

    Energy Technology Data Exchange (ETDEWEB)

    Anast, Kurt Roy [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-11-13

    This report examines and assesses the available systems and facilities considered for carrying out remediation activities on remediated nitrate salt (RNS) and unremediated nitrate salt (UNS) waste containers at Los Alamos National Laboratory (LANL). The assessment includes a review of the waste streams consisting of 60 RNS, 29 above-ground UNS, and 79 candidate below-ground UNS containers that may need remediation. The waste stream characteristics were examined along with the proposed treatment options identified in the Options Assessment Report . Two primary approaches were identified in the five candidate treatment options discussed in the Options Assessment Report: zeolite blending and cementation. Systems that could be used at LANL were examined for housing processing operations to remediate the RNS and UNS containers and for their viability to provide repackaging support for remaining LANL legacy waste.

  6. Engineering Options Assessment Report: Nitrate Salt Waste Stream Processing

    Energy Technology Data Exchange (ETDEWEB)

    Anast, Kurt Roy [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-11-18

    This report examines and assesses the available systems and facilities considered for carrying out remediation activities on remediated nitrate salt (RNS) and unremediated nitrate salt (UNS) waste containers at Los Alamos National Laboratory (LANL). The assessment includes a review of the waste streams consisting of 60 RNS, 29 aboveground UNS, and 79 candidate belowground UNS containers that may need remediation. The waste stream characteristics were examined along with the proposed treatment options identified in the Options Assessment Report . Two primary approaches were identified in the five candidate treatment options discussed in the Options Assessment Report: zeolite blending and cementation. Systems that could be used at LANL were examined for housing processing operations to remediate the RNS and UNS containers and for their viability to provide repackaging support for remaining LANL legacy waste.

  7. Investigation of Various LiCl Waste Salt Purification Technologies

    International Nuclear Information System (INIS)

    Yung-Zun Cho; Hee-Chul Yang; Han-Soo Lee; In-Tae Kim

    2008-01-01

    Various purification research of LiCl waste molten salt generated from electroreduction process were tested. The purification of the LiCl waste salt very important in a various aspects, where the purification means separation of cesium and strontium form LiCl salt melts. In this study, for the separation of cesium and strontium from LiCl salt melts, precipitant agent addition techniques such as sulfate and carbonate addition method and, as a new attempt, zone freezing technique for concentration of cesium and strontium elements was investigated. As a results of this research, only strontium was carbonated by reaction with Li 2 CO 3 (cesium did not react with Li 2 CO 3 ). In case of sulfate addition method, both cesium and strontium were converted into their sulfate that is Cs 2 S 2 O 6 and SrSO 4 and maximum sulfate efficiency of cesium and strontium were about 72% and 95%, respectively. Cesium and strontium involved in LiCl molten salt could be concentrated in the molten salt by using zone freezing method. (authors)

  8. Cerebral salt wasting following tuberculous meningoencephalitis in an infant

    Directory of Open Access Journals (Sweden)

    Syed Ahmed Zaki

    2012-01-01

    Full Text Available In patients with central nervous system disease, life-threatening hyponatremia can result from either the syndrome of inappropriate secretion of antidiuretic hormone or cerebral salt wasting. Clinical manifestations of the two conditions may be similar, but their pathogeneses and management protocols are different. Cerebral salt wasting syndrome is a disorder in which excessive natriuresis and hyponatremia occurs in patients with intracranial diseases. We report a 6-month-old girl with CSWS associated with tuberculous meningoencephalitis. She was diagnosed as having CSWS on the basis of hypovolemia, polyuria, natriuresis, and the relatively high level of fractional excretion of uric acid. Aggressive replacement of urine salt and water losses using 0.9% or 3% sodium chloride was done. Fludrocortisone was started at 0.1 mg twice daily on the seventh day of admission and was continued for 17 days.

  9. Concepts and Technologies for Radioactive Waste Disposal in Rock Salt

    Directory of Open Access Journals (Sweden)

    Wernt Brewitz

    2007-01-01

    Full Text Available In Germany, rock salt was selected to host a repository for radioactive waste because of its excellent mechanical properties. During 12 years of practical disposal operation in the Asse mine and 25 years of disposal in the disused former salt mine Morsleben, it was demonstrated that low-level wastes (LLW and intermediate-level wastes (ILW can be safely handled and economically disposed of in salt repositories without a great technical effort. LLW drums were stacked in old mining chambers by loading vehicles or emplaced by means of the dumping technique. Generally, the remaining voids were backfilled by crushed salt or brown coal filter ash. ILW were lowered into inaccessible chambers through a borehole from a loading station above using a remote control.Additionally, an in-situ solidification of liquid LLW was applied in the Morsleben mine. Concepts and techniques for the disposal of heat generating high-level waste (HLW are advanced as well. The feasibility of both borehole and drift disposal concepts have been proved by about 30 years of testing in the Asse mine. Since 1980s, several full-scale in-situ tests were conducted for simulating the borehole emplacement of vitrified HLW canisters and the drift emplacement of spent fuel in Pollux casks. Since 1979, the Gorleben salt dome has been investigated to prove its suitability to host the national final repository for all types of radioactive waste. The “Concept Repository Gorleben” disposal concepts and techniques for LLW and ILW are widely based on the successful test operations performed at Asse. Full-scale experiments including the development and testing of adequate transport and emplacement systems for HLW, however, are still pending. General discussions on the retrievability and the reversibility are going on.

  10. Salt Repository Project waste emplacement mode decision paper: Revison 1

    International Nuclear Information System (INIS)

    1987-08-01

    This paper provides a recommendation as to the mode of waste emplacement to be used as the current basis for site characterization activity for the Deaf Smith County, Texas, high level nuclear waste repository site. It also presents a plan for implementing the recommendation so as to provide a high level of confidence in the project's success. Since evaluations of high-level waste disposal in geologic repositories began in the 1950s, most studies emplacement in salt formations employed the vertical orientation for emplacing waste packages in boreholes in the floor of the underground facility. This orientation was used in trials at Project Salt Vault in the 1960s. The Waste Isolation Pilot Plant (WIPP) has recently settled on a combination of vertical and horizontal modes for various waste types. This paper analyzes the information available and develops a project position upon which to base current site characterization activities. The position recommended is that the SRP should continue to use the vertical waste emplacement mode as the reference design and to carry the horizontal mode as a ''passive'' alternative. This position was developed based upon the conclusions of a decision analysis, risk assessment, and cost/schedule impact assessment. 52 refs., 6 figs., 1 tab

  11. Novel waste printed circuit board recycling process with molten salt.

    Science.gov (United States)

    Riedewald, Frank; Sousa-Gallagher, Maria

    2015-01-01

    The objective of the method was to prove the concept of a novel waste PCBs recycling process which uses inert, stable molten salts as the direct heat transfer fluid and, simultaneously, uses this molten salt to separate the metal products in either liquid (solder, zinc, tin, lead, etc.) or solid (copper, gold, steel, palladium, etc.) form at the operating temperatures of 450-470 °C. The PCB recovery reactor is essentially a U-shaped reactor with the molten salt providing a continuous fluid, allowing molten salt access from different depths for metal recovery. A laboratory scale batch reactor was constructed using 316L as suitable construction material. For safety reasons, the inert, stable LiCl-KCl molten salts were used as direct heat transfer fluid. Recovered materials were washed with hot water to remove residual salt before metal recovery assessment. The impact of this work was to show metal separation using molten salts in one single unit, by using this novel reactor methodology. •The reactor is a U-shaped reactor filled with a continuous liquid with a sloped bottom representing a novel reactor concept.•This method uses large PCB pieces instead of shredded PCBs as the reactor volume is 2.2 L.•The treated PCBs can be removed via leg B while the process is on-going.

  12. Novel waste printed circuit board recycling process with molten salt

    Science.gov (United States)

    Riedewald, Frank; Sousa-Gallagher, Maria

    2015-01-01

    The objective of the method was to prove the concept of a novel waste PCBs recycling process which uses inert, stable molten salts as the direct heat transfer fluid and, simultaneously, uses this molten salt to separate the metal products in either liquid (solder, zinc, tin, lead, etc.) or solid (copper, gold, steel, palladium, etc.) form at the operating temperatures of 450–470 °C. The PCB recovery reactor is essentially a U-shaped reactor with the molten salt providing a continuous fluid, allowing molten salt access from different depths for metal recovery. A laboratory scale batch reactor was constructed using 316L as suitable construction material. For safety reasons, the inert, stable LiCl–KCl molten salts were used as direct heat transfer fluid. Recovered materials were washed with hot water to remove residual salt before metal recovery assessment. The impact of this work was to show metal separation using molten salts in one single unit, by using this novel reactor methodology. • The reactor is a U-shaped reactor filled with a continuous liquid with a sloped bottom representing a novel reactor concept. • This method uses large PCB pieces instead of shredded PCBs as the reactor volume is 2.2 L. • The treated PCBs can be removed via leg B while the process is on-going. PMID:26150977

  13. Backfill barriers for nuclear waste repositories in salt

    Energy Technology Data Exchange (ETDEWEB)

    Nowak, E J; Odoj, R; Merz, E [eds.

    1981-06-01

    Backfill materials were evaluated for containment of radionuclides, chemical modification of brine, and sensitivity to hydrothermal conditions. Experimental conditions were relevant to nuclear waste isolation in bedded salt. They were based on geologic conditions at the site of the Waste Isolation Pilot Plant (WIPP) in southeastern New Mexico, USA. Conclusions are: backfill mixtures surrounding the waste form and canister can provide a neutral or slightly acidic, potentially reducing environment, prevent convective aqueous flow, and act as an effective radionuclide migration barrier; bentonite is likely to remain hydrothermally stable but potentially sensitive to waste package interactions which could alter the pH, the ratio of dissolved ions, or the sorption properties of radionuclide species; effects of irradiation from high level waste should be investigated.

  14. Effect of indomethacin and salt depletion on renal proton MR imaging; An experimental study in the rat

    Energy Technology Data Exchange (ETDEWEB)

    Heyman, S.N.; Mammen, M. (Harvard Medical School, Boston, MA (United States). Charles A Dana Research Inst. Beth Israel Hospital, Boston, MA (United States))

    1991-11-01

    Blockade of the synthesis of vasodilating prostaglandins with non-steroidal anti-inflammatory drugs (NSAID) renders the renal medulla susceptible to hypoxic injury with reduced renal function, especially in clinical conditions characterized by volume depletion. Alterations in renal hemodynamics and urine production may effect renal MR imaging under these circumstances. We injected salt-depleted and control rats undergoing proton MR imaging with indomethacin 10 mg/kg. Indomethacin abolished the cortico-medullary T2-gradient and markedly diminished the overall renal signal in salt-depleted rats only. These changes, which progressed over a period of 40 min after indomethacin was injected, probably result from renal oligemia and decreased urine production, with an associated decrease in T2-values. We suggest that a history of consumption of non-steroidal anti-inflammatory drugs should be obtained and taken into account in the evaluation of renal proton MR imaging, especially in the presence of salt and volume depletion. (orig.).

  15. Salt-occluded zeolite waste forms: Crystal structures and transformability

    International Nuclear Information System (INIS)

    Richardson, J.W. Jr.

    1996-01-01

    Neutron diffraction studies of salt-occluded zeolite and zeolite/glass composite samples, simulating nuclear waste forms loaded with fission products, have revealed complex structures, with cations assuming the dual roles of charge compensation and occlusion (cluster formation). These clusters roughly fill the 6--8 angstrom diameter pores of the zeolites. Samples are prepared by equilibrating zeolite-A with complex molten Li, K, Cs, Sr, Ba, Y chloride salts, with compositions representative of anticipated waste systems. Samples prepared using zeolite 4A (which contains exclusively sodium cations) as starting material are observed to transform to sodalite, a denser aluminosilicate framework structure, while those prepared using zeolite 5A (sodium and calcium ions) more readily retain the zeolite-A structure. Because the sodalite framework pores are much smaller than those of zeolite-A, clusters are smaller and more rigorously confined, with a correspondingly lower capacity for waste containment. Details of the sodalite structures resulting from transformation of zeolite-A depend upon the precise composition of the original mixture. The enhanced resistance of salt-occluded zeolites prepared from zeolite 5A to sodalite transformation is thought to be related to differences in the complex chloride clusters present in these zeolite mixtures. Data relating processing conditions to resulting zeolite composition and structure can be used in the selection of processing parameters which lead to optimal waste forms

  16. [Salt intake and the progression of renal failure in patients with chronic kidney disease].

    Science.gov (United States)

    Amaha, Mayuko; Ohashi, Yasushi; Sakai, Ken; Aikawa, Atsushi; Mizuiri, Sonoo

    2010-01-01

    Salt intake not only elevates the levels of blood pressure, glomerular capillary pressure and proteinuria, but also increases oxidative stress within the renal cortex in animal models. We examined the effect of salt intake on the rate of renal function decline, urinary protein and oxidative stress in patients with chronic kidney disease (CKD). Clinical data including systolic blood pressure (SBP)and diastolic blood pressure (DBP), serum creatinine, uric acid, total cholesterol, triglyceride, urinary protein, salt intake, protein intake of non-diabetic CKD 53 patients were observed for one year. At the end of the observation period, we measured 8-hydroxydeoxy guanosine (8-OHdG) in spot urine. We calculated the slope of reciprocal serum creatinine as the rate of renal function decline (delta1/Cr). We then investigated the relationship between those clinical factors and delta1/Cr, and urinary 8-OHdG, and also selected clinical factors that significantly influence delta1/Cr and urinary 8-OHdG by stepwise multiple regression analysis. In addition, we investigated the gender difference in urinary 8-OHdG. Annual mean SBP and DBP of all patients were 121.5 +/- 9.3 mmHg and 72.5+/- 6.2 mmHg, respectively. delta1/Cr was negatively correlated with salt intake, urinary protein and urinary protein was a significant predictor of delta1/Cr in a multiple regression analysis. Salt intake was positively correlated with protein intake and urinary protein. Urinary 8-OHdG of all patients was positively correlated with urinary protein and it was a significant predictor. Urinary 8-OHdG of male patients was positively correlated with salt intake and was a significant predictor; in female patients, it was positively correlated with urinary protein and total cholesterol and these two factors were significant predictors. Salt intake increases urinary protein and promotes the progression of renal failure in CKD patients.

  17. Cellulose-containing Waste and Bituminized Salts

    International Nuclear Information System (INIS)

    Valcke, E.

    2005-01-01

    In Belgium, Medium-Level radioactive Waste (MLW) would be eventually disposed off in an underground repository in a geological formation such as the Boom Clay, which is studied as a reference host rock formation. MLW contains large quantities of non-radioactive chemicals that are released upon contact with pore water. It could be the case, for instance, for plutonium bearing cellulosic waste - such as paper tissues used to clean alpha glove boxes - issued from nuclear fuel fabrication (Belgonucleaire). At high pH, as in a disposal gallery backfilled with cement, the chemical degradation of cellulose will generate water-soluble products that may form strong complexes with actinides such as Am, Pu, Np, and U. This could lower the sorption of these elements onto the clay minerals, and hence increase their migration through the clay barrier. Another chemical perturbation could occur from the 3000 m 3 of so-called Eurobitum bituminised MLW, with precipitation sludges from the chemical treatment of spent nuclear fuel, and containing about 750 tons of NaNO 3 . The presence of NaNO 3 in this waste will give rise to several processes susceptible to affect the safety of the disposal system. Amongst others, it is necessary to verify that the swelling pressure of bitumen on the gallery wall and the osmotic pressure within the near-field are not too high to induce a fissuration of the host rock, leading to the formation of preferential migration pathways. The major objective of our work is to obtain a broad understanding of the different processes induced by the release of non-radioactive chemicals in the clay formation, to assess the chemical compatibility of different MLW forms with the clay

  18. Differential regulation of renal cyclooxygenase mRNA by dietary salt intake

    DEFF Research Database (Denmark)

    Jensen, B L; Kurtz, A

    1997-01-01

    RNA correlated directly with salt intake. We conclude that dietary salt intake influences renal cyclooxygenase mRNAs zone-specifically with opposite responses between cortex and medulla. Cortical COX II-mediated prostaglandin formation is probably important in low salt states whereas medullary COX I......Experiments were done to investigate the influence of dietary salt intake on renal cyclooxygenase (COX) I and II mRNA levels. To this end rats were fed either a low NaCl diet (LS; 0.02% NaCl wt/wt) or a high NaCl diet (HS diet; 4% NaCl wt/wt) for 5, 10 and 20 days. After 10 days Na excretion...... differed 760-fold, plasma renin activity and renin mRNA were increased eight- and threefold in LS compared to HS animals. Total renal COX I mRNA decreased 50% following the LS diet and did not change after the HS diet. Conversely, COX II mRNA declined after HS intake and transiently increased after salt...

  19. Waste package designs for disposal of high-level waste in salt formations

    International Nuclear Information System (INIS)

    Basham, S.J. Jr.; Carr, J.A.

    1984-01-01

    In the United States of America the selected method for disposal of radioactive waste is mined repositories located in suitable geohydrological settings. Currently four types of host rocks are under consideration: tuff, basalt, crystalline rock and salt. Development of waste package designs for incorporation in mined salt repositories is discussed. The three pertinent high-level waste forms are: spent fuel, as disassembled and close-packed fuel pins in a mild steel canister; commercial high-level waste (CHLW), as borosilicate glass in stainless-steel canisters; defence high-level waste (DHLW), as borosilicate glass in stainless-steel canisters. The canisters are production and handling items only. They have no planned long-term isolation function. Each waste form requires a different approach in package design. However, the general geometry and the materials of the three designs are identical. The selected waste package design is an overpack of low carbon steel with a welded closure. This container surrounds the waste forms. Studies to better define brine quantity and composition, radiation effects on the salt and brines, long-term corrosion behaviour of the low carbon steel, and the leaching behaviour of the spent fuel and borosilicate glass waste forms are continuing. (author)

  20. Hydrometallurgical treatment of plutonium. Bearing salt baths waste

    International Nuclear Information System (INIS)

    Bros, P.; Gozlan, J.P.; Lecomte, M.; Bourges, J.

    1993-01-01

    The salt flux issuing from the electrorefining of plutonium metal alloy in salt baths (KCI + NaCI) poses a difficult problem of the back-end alpha waste management. An alternative to the salt process promoted by Los Alamos Laboratory is to develop a hydrometallurgical treatment. A new process based on the electrochemistry technique in aqueous solution has been defined and tested successfully in the CEA. The diagram of the process exhibits two principal steps: in the head-end, a dissolution in HNO 3 medium accompanied with an electrolytic dechlorination leading to a quantitative elimination of chloride as CI 2 gas followed by its trapping one soda lime cartridge, a complete oxidative dissolution of the refractory Pu residues by electrogenerated Ag(II), in the back-end: the Pu and Am recoveries by chromatographic extractions. (authors). 10 figs., 9 refs

  1. Salt-induced changes in cardiac phosphoproteome in a rat model of chronic renal failure.

    Directory of Open Access Journals (Sweden)

    Zhengxiu Su

    Full Text Available Heart damage is widely present in patients with chronic kidney disease. Salt diet is the most important environmental factor affecting development of chronic renal failure and cardiovascular diseases. The proteins involved in chronic kidney disease -induced heart damage, especially their posttranslational modifications, remain largely unknown to date. Sprague-Dawley rats underwent 5/6 nephrectomy (chronic renal failure model or sham operation were treated for 2 weeks with a normal-(0.4% NaCl, or high-salt (4% NaCl diet. We employed TiO2 enrichment, iTRAQ labeling and liquid-chromatography tandem mass spectrometry strategy for phosphoproteomic profiling of left ventricular free walls in these animals. A total of 1724 unique phosphopeptides representing 2551 non-redundant phosphorylation sites corresponding to 763 phosphoproteins were identified. During normal salt feeding, 89 (54% phosphopeptides upregulated and 76 (46% phosphopeptides downregulated in chronic renal failure rats relative to sham rats. In chronic renal failure rats, high salt intake induced upregulation of 84 (49% phosphopeptides and downregulation of 88 (51% phosphopeptides. Database searches revealed that most of the identified phospholproteins were important signaling molecules such as protein kinases, receptors and phosphatases. These phospholproteins were involved in energy metabolism, cell communication, cell differentiation, cell death and other biological processes. The Search Tool for the Retrieval of Interacting Genes analysis revealed functional links among 15 significantly regulated phosphoproteins in chronic renal failure rats compared to sham group, and 23 altered phosphoproteins induced by high salt intake. The altered phosphorylation levels of two proteins involved in heart damage, lamin A and phospholamban were validated. Expression of the downstream genes of these two proteins, desmin and SERCA2a, were also analyzed.

  2. Radioactive waste disposal in the Gorleben salt deposit

    International Nuclear Information System (INIS)

    Gizycki, P. von

    1985-01-01

    In the opinion of five experts, the protective function of the overlying rock as a barrier has turned out to be questionable after borings and measurements carried through at Gorleben. Moreover, the results have also raised doubts about the geological safety of the salt deposit as a barrier in the long run. The geological multibarrier concept must be discarded. Not only critics, but also 3 advocates from the field of official research on radioactive waste disposal state their opinion. (DG) [de

  3. Performance analysis of conceptual waste package designs in salt repositories

    International Nuclear Information System (INIS)

    Jansen, G. Jr.; Raines, G.E.; Kircher, J.F.

    1984-01-01

    A performance analysis of commercial high-level waste and spent fuel conceptual package designs in reference repositories in three salt formations was conducted with the WAPPA waste package code. Expected conditions for temperature, stress, brine composition, radiation level, and brine flow rate were used as boundary conditions to compute expected corrosion of a thick-walled overpack of 1025 wrought steel. In all salt formations corrosion by low Mg salt-dissolution brines typical of intrusion scenarios was too slow to cause the package to fail for thousands of years after burial. In high Mg brines judged typical of thermally migrating brines in bedded salt formations, corrosion rates which would otherwise have caused the packages to fail within a few hundred years were limited by brine availability. All of the brine reaching the package was consumed by reaction with the iron in the overpack, thus preventing further corrosion. Uniform brine distribution over the package surface was an important factor in predicting long package lifetimes for the high Mg brines. 14 references, 15 figures

  4. Nuclear waste in sea or salt? No, wrong

    International Nuclear Information System (INIS)

    Damveld, H.; Van Duin, S.; Bannink, D.

    1994-04-01

    Eighteen years of successful action against ocean dumping and storage of nuclear waste in salt domes are reviewed for the Dutch situation. The aim of this book is to hand some support to those who want to act against trial borings, in particular the people living close to the most important salt domes in the Netherlands: Ternaard, Zuidwending, Pieterburen, Onstwedde, Winschoten, Schoonlo and Gasselte-Drouwen. In 1976 the Interdepartmental Commission on Nuclear Energy with its subcommission Radioactive Substances (ICK-RAS) was installed, along with a number of working groups, responsible for research. From 1978 onwards ocean dumping operations were accompanied by blockades and legal procedures, which led to a situation of the last dumping in 1982. The Dutch government then focused on nuclear waste storage in salt domes for which the OPLA research program was started. OPLA is the Dutch abbreviation for Storage on Land. The final report (phase 1 and 1a) of OPLA was published on 15 October 1993 as annex to the Dossier Nuclear Energy of the Dutch government. It has been decided that phase 1a is not followed by trial drillings, as planned before. Some critical remarks are made regarding the rounds of public participation and the notion of permanent retrievability of stored nuclear waste. Extensive use has been made of documentation from the Dutch government and parliament, and other literature and information sources

  5. Problems of the final storage of radioactive waste in salt formations

    International Nuclear Information System (INIS)

    Hofrichter, E.

    1977-01-01

    The geological conditions for the final storage of radioactive waste, the occurrence of salt formations, and the tectonics of salt domes are discussed. The safety of salt rocks, the impermeability of the rocks, and the thermal problems in the storage of high-activity waste are dealt with. Possibilities and preconditions of final storage in West Germany are discussed. (HPH) [de

  6. [Salt, renal function and high blood pressure--reflections on a current issue].

    Science.gov (United States)

    Aurell, Mattias

    2002-11-21

    The role of salt intake for blood pressure control has been discussed for a long time. A brief review is given of some pertinent physiological facts to explain this relationship and evolutionary aspects of renal function are emphasized. Salt intake is very high in the modern society, often as high as 15 g sodium chloride per 24 hours while 3-6 g may be more than enough to maintain an adequate salt balance. If the kidneys cannot cope with this severe sodium overload, blood pressure will rise. Therefore, the kidneys' ability to excrete sodium is a key factor and the salt excretion capacity is the kidneys' major barostatic function. As barostats, the kidneys control the blood pressure by ultimately determining the sodium excretion. Reducing sodium intake is, however, difficult as more than 50% of the intake is contained in the food we buy such as bread, sausages, canned food, chips and fast-food. Food products should therefore be "salt declared", but information on this aspect is generally lacking. If the population's salt intake could be reduced by 50%, the prevalence of hypertension will be much reduced, perhaps also by as much as 50%. The cost to society for treating hypertension would be reduced accordingly. Salt intake is also an important aspect of the overweight problem among today's youth. Salt and overweight impose great health risks later in life. Preventive measures in this area must be given high priority in future health care work.

  7. Salt creep design consideration for underground nuclear waste storage

    International Nuclear Information System (INIS)

    Li, W.T.; Wu, C.L.; Antonas, N.J.

    1983-01-01

    This paper summarizes the creep consideration in the design of nuclear waste storage facilities in salt, describes the non-linear analysis method for evaluating the design adequacy, and presents computational results for the current storage design. The application of rock mechanics instrumentation to assure the appropriateness of the design is discussed. It also describes the design evolution of such a facility, starting from the conceptual design, through the preliminary design, to the detailed design stage. The empirical design method, laboratory tests and numerical analyses, and the underground in situ tests have been incorporated in the design process to assure the stability of the underground openings, retrievability of waste during the operation phase and encapsulation of waste after decommissioning

  8. Trial storage of high-level waste in the Asse II salt mine

    International Nuclear Information System (INIS)

    1984-01-01

    This report covers a second phase of the work performed by GSF and KfK in the Asse II salt mine, with a view to disposal of radioactive waste in salt formations. New items of the research were geophysical investigations of the behaviour of heated salt and preparation of a trial storage in the Asse II salt mine

  9. Waste package reference conceptual designs for a repository in salt

    International Nuclear Information System (INIS)

    1986-02-01

    This report provides the reference conceptual waste package designs for the Office of Nuclear Waste Isolation to baseline these designs, thereby establishing the configuration and interface controls necessary, within the Civilian Radioactive Waste Management Program, formerly the National Waste Terminal Storage Program, to proceed in an orderly manner with preliminary design. Included are designs for the current reference defense high-level waste form from the Savannah River Plant, an optimized commercial high-level waste form, and spent fuel which has been disassembled and compacted into a circular bundle containing either 12 pressurized-water reactor or 30 boiling-water reactor assemblies. For compacted spent fuel, it appears economically attractive to standardize the waste package diameter for all fuel types. The reference waste packages consist of the containerized waste form, a low carbon steel overpack, and, after emplacement, a cover of salt. The overpack is a hollow cylinder with a flat head welded to each end. Its design thickness is the sum of the structural thickness required to resist the 15.4-MPa lithostatic pressure plus the corrosion allowance necessary to assure the required structural thickness will exist through the 1000-year containment period. Based on available data and completed analyses, the reference concepts described in this report satisfy all requirements of the US Department of Energy and the US Nuclear Regulatory Commission with reasonable assurance. In addition, sufficient design maturity exists to form a basis for preliminary design; these concepts can be brought under configuration control to serve as reference package designs. Development programs are identified that will be required to support these designs during the licensing process. 19 refs., 37 figs., 31 tabs

  10. Prostaglandin-E2 Mediated Increase in Calcium and Phosphate Excretion in a Mouse Model of Distal Nephron Salt Wasting.

    Directory of Open Access Journals (Sweden)

    Manoocher Soleimani

    Full Text Available Contribution of salt wasting and volume depletion to the pathogenesis of hypercalciuria and hyperphosphaturia is poorly understood. Pendrin/NCC double KO (pendrin/NCC-dKO mice display severe salt wasting under basal conditions and develop profound volume depletion, prerenal renal failure, and metabolic alkalosis and are growth retarded. Microscopic examination of the kidneys of pendrin/NCC-dKO mice revealed the presence of calcium phosphate deposits in the medullary collecting ducts, along with increased urinary calcium and phosphate excretion. Confirmatory studies revealed decreases in the expression levels of sodium phosphate transporter-2 isoforms a and c, increases in the expression of cytochrome p450 family 4a isotypes 12 a and b, as well as prostaglandin E synthase 1, and cyclooxygenases 1 and 2. Pendrin/NCC-dKO animals also had a significant increase in urinary prostaglandin E2 (PGE-2 and renal content of 20-hydroxyeicosatetraenoic acid (20-HETE levels. Pendrin/NCC-dKO animals exhibit reduced expression levels of the sodium/potassium/2chloride co-transporter 2 (NKCC2 in their medullary thick ascending limb. Further assessment of the renal expression of NKCC2 isoforms by quantitative real time PCR (qRT-PCR reveled that compared to WT mice, the expression of NKCC2 isotype F was significantly reduced in pendrin/NCC-dKO mice. Provision of a high salt diet to rectify volume depletion or inhibition of PGE-2 synthesis by indomethacin, but not inhibition of 20-HETE generation by HET0016, significantly improved hypercalciuria and salt wasting in pendrin/NCC dKO mice. Both high salt diet and indomethacin treatment also corrected the alterations in NKCC2 isotype expression in pendrin/NCC-dKO mice. We propose that severe salt wasting and volume depletion, irrespective of the primary originating nephron segment, can secondarily impair the reabsorption of salt and calcium in the thick ascending limb of Henle and/or proximal tubule, and reabsorption of

  11. Pyrolytic conversion of plastic and rubber waste to hydrocarbons with basic salt catalysts

    Science.gov (United States)

    Wingfield, Jr., Robert C.; Braslaw, Jacob; Gealer, Roy L.

    1985-01-01

    The invention relates to a process for improving the pyrolytic conversion of waste selected from rubber and plastic to low molecular weight olefinic materials by employing basis salt catalysts in the waste mixture. The salts comprise alkali or alkaline earth compounds, particularly sodium carbonate, in an amount of greater than about 1 weight percent based on the waste feed.

  12. Assessment of tectonic hazards to waste storage in interior-basin salt domes

    International Nuclear Information System (INIS)

    Kehle, R.

    1979-01-01

    Salt domes in the northern Gulf of Mexico may make ideal sites for storage of radioactive waste because the area is tectonically quiet. The stability of such salt domes and the tectonic activity are discussed

  13. High-Level Waste Salt Disposition Systems Engineering Team Final Report, Volumes I, II, and III

    International Nuclear Information System (INIS)

    Piccolo, S.F.

    1999-01-01

    This report describes the process used and results obtained by the High Level Waste Salt Disposition Systems Engineering Team to select a primary and backup alternative salt disposition method for the Savannah River Site

  14. Stabilization Using Phosphate Bonded Ceramics. Salt Containing Mixed Waste Treatment. Mixed Waste Focus Area. OST Reference No. 117

    International Nuclear Information System (INIS)

    1999-01-01

    Throughout the Department of Energy (DOE) complex there are large inventories of homogeneous mixed waste solids, such as wastewater treatment residues, fly ashes, and sludges that contain relatively high concentrations (greater than 15% by weight) of salts. The inherent solubility of salts (e.g., nitrates, chlorides, and sulfates) makes traditional treatment of these waste streams difficult, expensive, and challenging. One alternative is low-temperature stabilization by chemically bonded phosphate ceramics (CBPCs). The process involves reacting magnesium oxide with monopotassium phosphate with the salt waste to produce a dense monolith. The ceramic makes a strong environmental barrier, and the metals are converted to insoluble, low-leaching phosphate salts. The process has been tested on a variety of surrogates and actual mixed waste streams, including soils, wastewater, flyashes, and crushed debris. It has also been demonstrated at scales ranging from 5 to 55 gallons. In some applications, the CBPC technology provides higher waste loadings and a more durable salt waste form than the baseline method of cementitious grouting. Waste form test specimens were subjected to a variety of performance tests. Results of waste form performance testing concluded that CBPC forms made with salt wastes meet or exceed both RCRA and recommended Nuclear Regulatory Commission (NRC) low-level waste (LLW) disposal criteria. Application of a polymer coating to the CBPC may decrease the leaching of salt anions, but continued waste form evaluations are needed to fully assess the deteriorating effects of this leaching, if any, over time.

  15. Deep geologic disposal of mixed waste in bedded salt: The Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    Rempe, N.T.

    1993-01-01

    Mixed waste (i.e., waste that contains both chemically hazardous and radioactive components) poses a moral, political, and technical challenge to present and future generations. But an international consensus is emerging that harmful byproducts and residues can be permanently isolated from the biosphere in a safe and environmentally responsible manner by deep geologic disposal. To investigate and demonstrate such disposal for transuranic mixed waste, derived from defense-related activities, the US Department of Energy has prepared the Waste Isolation Pilot Plant (WIPP) near Carlsbad, New Mexico. This research and development facility was excavated approximately at the center of a 600 m thick sequence of salt (halite) beds, 655 m below the surface. Proof of the long-term tectonic and hydrological stability of the region is supplied by the fact that these salt beds have remained essentially undisturbed since they were deposited during the Late Permian age, approximately 225 million years ago. Plutonium-239, the main radioactive component of transuranic mixed waste, has a half-life of 24,500 years. Even ten half-lives of this isotope - amounting to about a quarter million years, the time during which its activity will decline to background level represent only 0.11 percent of the history of the repository medium. Therefore, deep geologic disposal of transuranic mixed waste in Permian bedded salt appears eminently feasible

  16. Altered regulation of renal sodium transporters in salt-sensitive hypertensive rats induced by uninephrectomy.

    Science.gov (United States)

    Jung, Ji Yong; Lee, Jay Wook; Kim, Sejoong; Jung, Eun Sook; Jang, Hye Ryoun; Han, Jin Suk; Joo, Kwon Wook

    2009-12-01

    Uninephrectomy (uNx) in young rats causes salt-sensitive hypertension (SSH). Alterations of sodium handling in residual nephrons may play a role in the pathogenesis. Therefore, we evaluated the adaptive alterations of renal sodium transporters according to salt intake in uNx-SSH rats. uNx or sham operations were performed in male Sprague-Dawley rats, and normal-salt diet was fed for 4 weeks. Four experimental groups were used: sham-operated rats raised on a high-salt diet for 2 weeks (CHH) or on a low-salt diet for 1 week after 1 week's high-salt diet (CHL) and uNx rats fed on the same diet (NHH, NHL) as the sham-operated rats were fed. Expression of major renal sodium transporters were determined by semiquantitative immunoblotting. Systolic blood pressure was increased in NHH and NHL groups, compared with CHH and CHL, respectively. Protein abundances of Na(+)/K(+)/2Cl(-) cotransporter (NKCC2) and Na(+)/Cl(-) cotransporter (NCC) in the CHH group were lower than the CHL group. Expression of epithelial sodium channel (ENaC)-γ increased in the CHH group. In contrast, expressions of NKCC2 and NCC in the NHH group didn't show any significant alterations, compared to the NHL group. Expressions of ENaC-α and ENaC-β in the NHH group were higher than the CHH group. Adaptive alterations of NKCC2 and NCC to changes of salt intake were different in the uNx group, and changes in ENaC-α and ENaC-β were also different. These altered regulations of sodium transporters may be involved in the pathogenesis of SSH in the uNx rat model.

  17. Hypokalemic salt-losing tubulopathy with chronic renal failure and sensorineural deafness.

    Science.gov (United States)

    Jeck, N; Reinalter, S C; Henne, T; Marg, W; Mallmann, R; Pasel, K; Vollmer, M; Klaus, G; Leonhardt, A; Seyberth, H W; Konrad, M

    2001-07-01

    To characterize a rare inherited hypokalemic salt-losing tubulopathy with linkage to chromosome 1p31. We conducted a retrospective analysis of the clinical data for 7 patients in whom cosegregation of the disease with chromosome 1p31 had been demonstrated. In addition, in 1 kindred, prenatal diagnosis in the second child was established, allowing a prospective clinical evaluation. Clinical presentation of the patients was homogeneous and included premature birth attributable to polyhydramnios, severe renal salt loss, normotensive hyperreninemia, hypokalemic alkalosis, and excessive hyperprostaglandin E-uria, which suggested the diagnosis of hyperprostaglandin E syndrome/antenatal Bartter syndrome. However, the response to indomethacin was only poor, accounting for a more severe variant of the disease. The patients invariably developed chronic renal failure. The majority had extreme growth retardation, and motor development was markedly delayed. In addition, all patients turned out to be deaf. The hypokalemic salt-losing tubulopathy with chronic renal failure and sensorineural deafness represents not only genetically but also clinically a disease entity distinct from hyperprostaglandin E syndrome/antenatal Bartter syndrome. A pleiotropic effect of a single gene defect is most likely causative for syndromic hearing loss.

  18. Cerebral salt wasting: a report of three cases

    International Nuclear Information System (INIS)

    Younas, H.; Sabir, O.; Tarif, N.

    2015-01-01

    Hyponatremia secondary to the Syndrome of Inappropriate Anti-Diuretic Hormone (SIADH) secretion is commonly observed in patients with various neurological disorders. Cerebral Salt Wasting (CSW) resulting in hyponatremia is also an infrequent occurrence in some patients with neurological disorders. Confusion in differentiating CSW from SIADH may arise since both results in similar electrolyte disturbances. Herein, we report three cases of CSW with intracranial afflictions. CSW was diagnosed on the basis of fractional excretion of urinary sodium and uric acid along with extremely low serum uric acid. Improvements in serum sodium levels after saline hydration and fludrocortisone administration further supported the diagnosis. (author)

  19. Areal thermal loading recommendations for nuclear waste repositories in salt

    International Nuclear Information System (INIS)

    Russell, J.E.

    1979-06-01

    This document gives a wider understanding of the history of the recommended thermal loadings in salt for both high-level waste (HLW) from fresh UO 2 -fueled, light-water reactors (LWR) with no recycle and spent unreprocessed fuel (SURF) from LWRs. Aspects of the current recommendations that need further study are identified. Finally, an interim set of design thermal-loading recommendations are given that have a common rationale of satisfying performance limits within our current state of knowledge. These recommendations are made on a generic rather than a site-specific basis. 11 figures, 5 tables

  20. Modeling of Sulfate Double-Salt in Nuclear Wastes

    International Nuclear Information System (INIS)

    Toghiani, B.; Lindner, J.S.; Weber, C.F.; Hunt, R.D.

    2000-01-01

    The Environmental Simulation Program (ESP) continues to adequately predict the solubility of most key chemical systems in the Hanford tank waste. For example, the ESP predictions were in fair agreement with the solubility experiments for the fluoride-phosphate system, although ESP probably underestimates the aqueous amounts. Due to the importance of this system in the formation of pipeline plugs, additional experiments have been made at elevated temperatures, and improvements to the ESP database will be made. ESP encountered problems with sulfate systems because the Public database for ESP does not include anhydrous sodium sulfate in mixed solutions below 32.4 C. This limitation leads to convergence problems and to spurious predictions of solubility near the transition point with sodium sulfate decahydrate when other salts such as sodium nitrate are present. However, ESP was able to make reasonable solubility predictions with a corrected database, demonstrating the need to validate and document the various databases that can be used by ESP. Even though ESP does not include the sulfate-nitrate double salt, this omission does not appear to be a major problem. The solubility predictions with and without the sulfate-nitrate double salt are comparable. In sharp contrast, the sulfate-fluoride double salt is included, but ESP still underestimates solubility in some cases. This problem can misrepresent the ionic strength of the solution, which is an important factor in the formation of pipeline plugs. Solubility tests on the sulfate-fluoride system are planned to provide additional data at higher temperatures and in caustic solutions. These results will be used to improve the range and accuracy of ESP predictions. ESP will continue to provide important predictions for waste processing operations while being evaluated and improved. For example, ESP will be used to determine the amount of water for the saltcake dissolution efforts at Hanford. When ESP underestimates the

  1. Identifying suitable piercement salt domes for nuclear waste storage sites

    International Nuclear Information System (INIS)

    Kehle, R.; e.

    1980-08-01

    Piercement salt domes of the northern interior salt basins of the Gulf of Mexico are being considered as permanent storage sites for both nuclear and chemically toxic wastes. The suitable domes are stable and inactive, having reached their final evolutionary configuration at least 30 million years ago. They are buried to depths far below the level to which erosion will penetrate during the prescribed storage period and are not subject to possible future reactivation. The salt cores of these domes are themselves impermeable, permitting neither the entry nor exit of ground water or other unwanted materials. In part, a stable dome may be recognized by its present geometric configuration, but conclusive proof depends on establishing its evolutionary state. The evolutionary state of a dome is obtained by reconstructing the growth history of the dome as revealed by the configuration of sedimentary strata in a large area (commonly 3,000 square miles or more) surrounding the dome. A high quality, multifold CDP reflection seismic profile across a candidate dome will provide much of the necessary information when integrated with available subsurface control. Additional seismic profiles may be required to confirm an apparent configuration of the surrounding strata and an interpreted evolutionary history. High frequency seismic data collected in the near vicinity of a dome are also needed as a supplement to the CDP data to permit accurate depiction of the configuration of shallow strata. Such data must be tied to shallow drill hole control to confirm the geologic age at which dome growth ceased. If it is determined that a dome reached a terminal configuration many millions of years ago, such a dome is incapable of reactivation and thus constitutes a stable storage site for nuclear wastes

  2. Maternal diet during gestation and lactation modifies the severity of salt-induced hypertension and renal injury in Dahl salt-sensitive rats.

    Science.gov (United States)

    Geurts, Aron M; Mattson, David L; Liu, Pengyuan; Cabacungan, Erwin; Skelton, Meredith M; Kurth, Theresa M; Yang, Chun; Endres, Bradley T; Klotz, Jason; Liang, Mingyu; Cowley, Allen W

    2015-02-01

    Environmental exposure of parents or early in life may affect disease development in adults. We found that hypertension and renal injury induced by a high-salt diet were substantially attenuated in Dahl SS/JrHsdMcwiCrl (SS/Crl) rats that had been maintained for many generations on the grain-based 5L2F diet compared with SS/JrHsdMcwi rats (SS/Mcw) maintained on the casein-based AIN-76A diet (mean arterial pressure, 116±9 versus 154±25 mm Hg; urinary albumin excretion, 23±12 versus 170±80 mg/d). RNAseq analysis of the renal outer medulla identified 129 and 82 genes responding to a high-salt diet uniquely in SS/Mcw and SS/Crl rats, respectively, along with minor genetic differences between the SS substrains. The 129 genes responding to salt in the SS/Mcw strain included numerous genes with homologs associated with hypertension, cardiovascular disease, or renal disease in human. To narrow the critical window of exposure, we performed embryo-transfer experiments in which single-cell embryos from 1 colony (SS/Mcw or SS/Crl) were transferred to surrogate mothers from the other colony, with parents and surrogate mothers maintained on their respective original diet. All offspring were fed the AIN-76A diet after weaning. Salt-induced hypertension and renal injury were substantially exacerbated in rats developed from SS/Crl embryos transferred to SS/Mcw surrogate mothers. Conversely, salt-induced hypertension and renal injury were significantly attenuated in rats developed from SS/Mcw embryos transferred to SS/Crl surrogate mothers. Together, the data suggest that maternal diet during the gestational-lactational period has substantial effects on the development of salt-induced hypertension and renal injury in adult SS rats. © 2014 American Heart Association, Inc.

  3. Establishment of cooperation basis of joint research on the mixed waste molten salt oxidation technology

    International Nuclear Information System (INIS)

    Yang, Hee Chul; Cho, Y. J.; Kim, J. H.; Yoo, J. H.; Yun, H. C.; Lee, D. G.

    2005-08-01

    Molten salt oxidation, MSO for short, is a robust technology that can effectively treat mixed waste (radioactive waste including hazardous metals or organics). It can safely and economically treat the difficult wastes such as not-easily destroyable toxic organic waste, medical waste, chemical warfare and energetic materials such as propellant and explosives, all of which are not easily treated by an incinerator or other currently existing thermal treatment system. Therefore, molten salt oxidation technology should be developed and utilized to treat a lot of niche waste stored in the nuclear and environmental industries. So, if we put the MSO technology to practical use by Korea-Vietnam joint research, we can reduce R and D fund for MSO technology by ourselves and we can expect an export of the outcome of nuclear R and D in Korea. For Establishment of cooperation basis of joint research concerning molten salt oxidation technology between KOREA and VIETNAM, in this research, We invited two Vietnamese researchers and we introduced our experimental scale molten salt oxidation system in order to let them understand molten salt oxidation technology. We also visited Viet man and we consulted about molten salt oxidation process. We held seminar on the mixed waste molten salt oxidation technology, discussed on the joint research on the mixed waste molten salt oxidation technology and finally we wrote MOU for joint research

  4. Establishment of cooperation basis of joint research on the mixed waste molten salt oxidation technology

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Hee Chul; Cho, Y. J.; Kim, J. H.; Yoo, J. H.; Yun, H. C.; Lee, D. G

    2005-08-01

    Molten salt oxidation, MSO for short, is a robust technology that can effectively treat mixed waste (radioactive waste including hazardous metals or organics). It can safely and economically treat the difficult wastes such as not-easily destroyable toxic organic waste, medical waste, chemical warfare and energetic materials such as propellant and explosives, all of which are not easily treated by an incinerator or other currently existing thermal treatment system. Therefore, molten salt oxidation technology should be developed and utilized to treat a lot of niche waste stored in the nuclear and environmental industries. So, if we put the MSO technology to practical use by Korea-Vietnam joint research, we can reduce R and D fund for MSO technology by ourselves and we can expect an export of the outcome of nuclear R and D in Korea. For Establishment of cooperation basis of joint research concerning molten salt oxidation technology between KOREA and VIETNAM, in this research, We invited two Vietnamese researchers and we introduced our experimental scale molten salt oxidation system in order to let them understand molten salt oxidation technology. We also visited Viet man and we consulted about molten salt oxidation process. We held seminar on the mixed waste molten salt oxidation technology, discussed on the joint research on the mixed waste molten salt oxidation technology and finally we wrote MOU for joint research.

  5. Treatment of waste salt from the advanced spent fuel conditioning process (I): characterization of Zeolite A in Molten LiCl Salt

    International Nuclear Information System (INIS)

    Kim, Jeong Guk; Lee, Jae Hee; Yoo, Jae Hyung; Kim, Joon Hyung

    2004-01-01

    The oxide fuel reduction process based on the electrochemical method (Advanced spent fuel Conditioning Process; ACP) and the long-lived radioactive nuclides partitioning process based on electro-refining process, which are being developed ay the Korea Atomic Energy Research Institute (KAERI), are to generate two types of molten salt wastes such as LiCl salt and LiCl-KCl eutectic salt, respectively. These waste salts must meet some criteria for disposal. A conditioning process for LiCl salt waste from ACP has been developed using zeolite A. This treatment process of waste salt using zeolite A was first developed by US ANL (Argonne National Laboratory) for LiCl-KCl eutectic salt waste from an electro-refining process of EBR (Experimental Breeder Reactor)-II spent fuel. This process has been developed recently, and a ceramic waste form (CWF) is produced in demonstration-scale V-mixer (50 kg/batch). However, ANL process is different from KAERI treatment process in waste salt, the former is LiCl-KCl eutectic salt and the latter is LiCl salt. Because of melting point, the immobilization of eutectic salt is carried out at about 770 K, whereas LiCl salt at around 920 K. Such difference has an effect on properties of immobilization media, zeolite A. Here, zeolite A in high-temperature (923 K) molten LiCl salt was characterized by XRD, Ion-exchange, etc., and evaluated if a promising media or not

  6. Method to synthesize dense crystallized sodalite pellet for immobilizing halide salt radioactive waste

    International Nuclear Information System (INIS)

    Koyama, Tadafumi.

    1994-01-01

    A method is described for immobilizing waste chloride salts containing radionuclides such as cesium and strontium and hazardous materials such as barium. A sodalite intermediate is prepared by mixing appropriate amounts of silica, alumina and sodium hydroxide with respect to sodalite and heating the mixture to form the sodalite intermediate and water. Heating is continued to drive off the water to form a water-free intermediate. The water-free intermediate is mixed with either waste salt or waste salt which has been contacted with zeolite to concentrate the radionuclides and hazardous material. The waste salt-intermediate mixture is then compacted and heated under conditions of heat and pressure to form sodalite with the waste salt, radionuclides and hazardous material trapped within the sodalite cage structure. This provides a final product having excellent leach resistant capabilities

  7. Interim performance specifications for conceptual waste-package designs for geologic isolation in salt repositories

    International Nuclear Information System (INIS)

    1983-06-01

    The interim performance specifications and data requirements presented apply to conceptual waste package designs for all waste forms which will be isolated in salt geologic repositories. The waste package performance specifications and data requirements respond to the waste package performance criteria. Subject areas treated include: containment and controlled release, operational period safety, criticality control, identification, and waste package performance testing requirements. This document was generated for use in the development of conceptual waste package designs in salt. It will be revised as additional data, analyses, and regulatory requirements become available

  8. Experimental results on salt concrete for barrier elements made of salt concrete in a repository for radioactive waste in a salt mine

    International Nuclear Information System (INIS)

    Gutsch, Alex-W.; Preuss, Juergen; Mauke, Ralf

    2012-01-01

    The Bartensleben rock salt mine in Germany was used as a repository for low and intermediate level radioactive waste from 1971 to 1991 and from 1994 to 1998. The repository with an overall volume of about 6 million m 3 has to be closed. Salt concrete is used for the refill of the voids of the repository. The concrete mixtures contain crushed salt instead of natural aggregates as the void filling material should be as similar to the salt rock as possible. Very high requirements regarding low heat development and little or even no cracking during concrete hardening had to be fulfilled even for the barrier elements made from salt concrete which separate the radioactive waste from the environment. Requirements for the salt concrete were set up with regard to the fluidity of the fresh concrete during the hardening process and its durability. In the view of a comprehensive numerical calculations of the temperature development and thermal stresses in the massive salt concrete elements of the backfill of the voids, experimental results for material properties of the salt concrete are presented: mixture of the salt concrete, thermodynamic properties (adiabatic heat release, thermal dilatation, thermal conductivity and heat capacity), mechanical short term properties, creep (under tension, under compression), autogenous shrinkage

  9. Container materials for isolation of radioactive waste in salt

    International Nuclear Information System (INIS)

    Streicher, M.A.; Andrews, A.

    1987-10-01

    The workshop reviewed the extensive data on the corrosion resistance of low-carbon steel in simulated salt repository environments, determined whether these data were sufficient to recommend low-carbon steel for fabrication of the container, and assessed the suitability of other materials under consideration in the SRP. The panelists determined the need for testing and research programs, recommended experimental approaches, and recommended materials based on existing technology. On the first day of the workshop, presentations were made on waste package requirements; the expected corrosion environment; degradation processes, including a review of data from corrosion tests on carbon steel; and rationales for container design and materials, modeling studies, and planned future work. The second day was devoted to a panel caucus, presentation of workshop findings, and open discussion. 76 refs., 2 figs., 3 tabs

  10. Test plan for immobilization of salt-containing surrogate mixed wastes using polyester resins

    International Nuclear Information System (INIS)

    Biyani, R.K.; Douglas, J.C.; Hendrickson, D.W.

    1997-01-01

    Past operations at many Department of Energy (DOE) sites have resulted in the generation of several waste streams with high salt content. These wastes contain listed and characteristic hazardous constituents and are radioactive. The salts contained in the wastes are primarily chloride, sulfate, nitrate, metal oxides, and hydroxides. DOE has placed these types of wastes under the purview of the Mixed Waste Focus Area (MWFA). The MWFA has been tasked with developing and facilitating the implementation of technologies to treat these wastes in support of customer needs and requirements. The MWFA has developed a Technology Development Requirements Document (TDRD), which specifies performance requirements for technology owners and developers to use as a framework in developing effective waste treatment solutions. This project will demonstrate the use of polyester resins in encapsulating and solidifying DOE's mixed wastes containing salts, as an alternative to conventional and other emerging immobilization technologies

  11. Test plan for immobilization of salt-containing surrogate mixed wastes using polyester resins

    Energy Technology Data Exchange (ETDEWEB)

    Biyani, R.K.; Douglas, J.C.; Hendrickson, D.W.

    1997-07-07

    Past operations at many Department of Energy (DOE) sites have resulted in the generation of several waste streams with high salt content. These wastes contain listed and characteristic hazardous constituents and are radioactive. The salts contained in the wastes are primarily chloride, sulfate, nitrate, metal oxides, and hydroxides. DOE has placed these types of wastes under the purview of the Mixed Waste Focus Area (MWFA). The MWFA has been tasked with developing and facilitating the implementation of technologies to treat these wastes in support of customer needs and requirements. The MWFA has developed a Technology Development Requirements Document (TDRD), which specifies performance requirements for technology owners and developers to use as a framework in developing effective waste treatment solutions. This project will demonstrate the use of polyester resins in encapsulating and solidifying DOE`s mixed wastes containing salts, as an alternative to conventional and other emerging immobilization technologies.

  12. Electrodialysis-based separation process for salt recovery and recycling from waste water

    Science.gov (United States)

    Tsai, Shih-Perng

    1997-01-01

    A method for recovering salt from a process stream containing organic contaminants is provided, comprising directing the waste stream to a desalting electrodialysis unit so as to create a concentrated and purified salt permeate and an organic contaminants containing stream, and contacting said concentrated salt permeate to a water-splitting electrodialysis unit so as to convert the salt to its corresponding base and acid.

  13. Renal Effects and Underlying Molecular Mechanisms of Long-Term Salt Content Diets in Spontaneously Hypertensive Rats

    Science.gov (United States)

    Berger, Rebeca Caldeira Machado; Vassallo, Paula Frizera; Crajoinas, Renato de Oliveira; Oliveira, Marilene Luzia; Martins, Flávia Letícia; Nogueira, Breno Valentim; Motta-Santos, Daisy; Araújo, Isabella Binotti; Forechi, Ludimila; Girardi, Adriana Castello Costa; Santos, Robson Augusto Souza; Mill, José Geraldo

    2015-01-01

    Several evidences have shown that salt excess is an important determinant of cardiovascular and renal derangement in hypertension. The present study aimed to investigate the renal effects of chronic high or low salt intake in the context of hypertension and to elucidate the molecular mechanisms underlying such effects. To this end, newly weaned male SHR were fed with diets only differing in NaCl content: normal salt (NS: 0.3%), low salt (LS: 0.03%), and high salt diet (HS: 3%) until 7 months of age. Analysis of renal function, morphology, and evaluation of the expression of the main molecular components involved in the renal handling of albumin, including podocyte slit-diaphragm proteins and proximal tubule endocytic receptors were performed. The relationship between diets and the balance of the renal angiotensin-converting enzyme (ACE) and ACE2 enzymes was also examined. HS produced glomerular hypertrophy and decreased ACE2 and nephrin expressions, loss of morphological integrity of the podocyte processes, and increased proteinuria, characterized by loss of albumin and high molecular weight proteins. Conversely, severe hypertension was attenuated and renal dysfunction was prevented by LS since proteinuria was much lower than in the NS SHRs. This was associated with a decrease in kidney ACE/ACE2 protein and activity ratio and increased cubilin renal expression. Taken together, these results suggest that LS attenuates hypertension progression in SHRs and preserves renal function. The mechanisms partially explaining these findings include modulation of the intrarenal ACE/ACE2 balance and the increased cubilin expression. Importantly, HS worsens hypertensive kidney injury and decreases the expression nephrin, a key component of the slit diaphragm. PMID:26495970

  14. Renal Effects and Underlying Molecular Mechanisms of Long-Term Salt Content Diets in Spontaneously Hypertensive Rats.

    Directory of Open Access Journals (Sweden)

    Rebeca Caldeira Machado Berger

    Full Text Available Several evidences have shown that salt excess is an important determinant of cardiovascular and renal derangement in hypertension. The present study aimed to investigate the renal effects of chronic high or low salt intake in the context of hypertension and to elucidate the molecular mechanisms underlying such effects. To this end, newly weaned male SHR were fed with diets only differing in NaCl content: normal salt (NS: 0.3%, low salt (LS: 0.03%, and high salt diet (HS: 3% until 7 months of age. Analysis of renal function, morphology, and evaluation of the expression of the main molecular components involved in the renal handling of albumin, including podocyte slit-diaphragm proteins and proximal tubule endocytic receptors were performed. The relationship between diets and the balance of the renal angiotensin-converting enzyme (ACE and ACE2 enzymes was also examined. HS produced glomerular hypertrophy and decreased ACE2 and nephrin expressions, loss of morphological integrity of the podocyte processes, and increased proteinuria, characterized by loss of albumin and high molecular weight proteins. Conversely, severe hypertension was attenuated and renal dysfunction was prevented by LS since proteinuria was much lower than in the NS SHRs. This was associated with a decrease in kidney ACE/ACE2 protein and activity ratio and increased cubilin renal expression. Taken together, these results suggest that LS attenuates hypertension progression in SHRs and preserves renal function. The mechanisms partially explaining these findings include modulation of the intrarenal ACE/ACE2 balance and the increased cubilin expression. Importantly, HS worsens hypertensive kidney injury and decreases the expression nephrin, a key component of the slit diaphragm.

  15. Expedited demonstration of molten salt mixed waste treatment technology. Final report

    International Nuclear Information System (INIS)

    1995-01-01

    This final report discusses the molten salt mixed waste project in terms of the various subtasks established. Subtask 1: Carbon monoxide emissions; Establish a salt recycle schedule and/or a strategy for off-gas control for MWMF that keeps carbon monoxide emission below 100 ppm on an hourly averaged basis. Subtask 2: Salt melt viscosity; Experiments are conducted to determine salt viscosity as a function of ash composition, ash concentration, temperature, and time. Subtask 3: Determine that the amount of sodium carbonate entrained in the off-gas is minimal, and that any deposited salt can easily be removed form the piping using a soot blower or other means. Subtask 4: The provision of at least one final waste form that meets the waste acceptance criteria of a landfill that will take the waste. This report discusses the progress made in each of these areas

  16. Polyethylene encapsulatin of nitrate salt wastes: Waste form stability, process scale-up, and economics

    International Nuclear Information System (INIS)

    Kalb, P.D.; Heiser, J.H. III; Colombo, P.

    1991-07-01

    A polyethylene encapsulation system for treatment of low-level radioactive, hazardous, and mixed wastes has been developed at Brookhaven National Laboratory. Polyethylene has several advantages compared with conventional solidification/stabilization materials such as hydraulic cements. Waste can be encapsulated with greater efficiency and with better waste form performance than is possible with hydraulic cement. The properties of polyethylene relevant to its long-term durability in storage and disposal environments are reviewed. Response to specific potential failure mechanisms including biodegradation, radiation, chemical attack, flammability, environmental stress cracking, and photodegradation are examined. These data are supported by results from extensive waste form performance testing including compressive yield strength, water immersion, thermal cycling, leachability of radioactive and hazardous species, irradiation, biodegradation, and flammability. The bench-scale process has been successfully tested for application with a number of specific ''problem'' waste streams. Quality assurance and performance testing of the resulting waste form confirmed scale-up feasibility. Use of this system at Rocky Flats Plant can result in over 70% fewer drums processed and shipped for disposal, compared with optimal cement formulations. Based on the current Rocky Flats production of nitrate salt per year, polyethylene encapsulation can yield an estimated annual savings between $1.5 million and $2.7 million, compared with conventional hydraulic cement systems. 72 refs., 23 figs., 16 tabs

  17. Low disposal of radioactive wastes in salt formations of the Federal Republic of Germaany

    International Nuclear Information System (INIS)

    Albrecht, E.

    1980-01-01

    The salt formations of northern Europe are generally suitable for the storage of radioactive wastes because the region is largely free from earthquakes and the salt formations known as diapires provide effective hydrological sealing. The Federal Republic of Germany employed the Asse Salt Mine of Lower Saxony for research in waste storage. More recently, exploratory work has begun on the construction of a large recycling and disposal plant at the Gorleben salt dome. The geology, hydrology, rock mechanics, and seismicity of the two sites are briefly discussed, including a discussion of experiences gained so far from the Asse site. 11 refs

  18. R and D activities on the management of waste chloride salts in KAERI

    International Nuclear Information System (INIS)

    In-Tae, Kim; Hwan-Seo, Park; Jeong-Gook, Kim; Hee-Chul, Yang; Yong-Joon, Cho; Eung-Ho Kim

    2007-01-01

    Full text of publication follows. Electrochemical treatment of spent oxide fuels has been intensively studied in KAERI to reduce the volume, heat load and radiotoxicity of high-level wastes. It consists of an electrolytic reduction process to convert the oxide fuel into a metallic form and an electro-refining process to separate TRU elements from the electro-reduced metal ingot. Two types of waste salts are expected to generate from the electrochemical pyro-processes, that is, LiCl salt from the reduction process and LiCl+KCl eutectic salt form the refining process. The R and D strategy of the waste salt management in KAERI can be categorized into two parts: 1) enhancement of safety by the stabilisation/solidification of waste salt that is to be finally disposed of and 2) reduction of the waste generation by the regeneration/recycle of the spent salt after removal of radionuclides in it. A sol-gel technique and a zeolite occlusion technique are under development to stabilize the waste salt. The LiCl salt is stabilised by a low-temperature sol-gel process and then the gel product is solidified into a ceramic-like waste form with an addition of glass frit. Another method uses Zeolite-4A to occlude the LiCl salt into its cage and adsorption site to immobilize the radionuclides. The product, salt-occluded zeolite, is fabricated into another type of a ceramic waste form. For the regeneration and recycle of the spent salt, the radionuclides in the salt are removed by a zeolite process for the LiCl salt and by an oxidation/distillation process for the eutectic salt. The target nuclides to be removed in each process are Cs/Sr and rare earth (RE) elements, respectively. In the oxidation/ distillation process, the rare earth chloride nuclides are oxidised by an oxygen sparging method, and the products are precipitated in the form of oxide or oxychloride REs. After separation of the RE elements from the precipitates by distillation, the refined spent salt with a low content

  19. Application of molten salt oxidation for the minimization and recovery of plutonium-238 contaminated wastes

    International Nuclear Information System (INIS)

    Wishau, R.; Ramsey, K.B.; Montoya, A.

    1998-01-01

    This paper presents the technical and economic feasibility of molten salt oxidation technology as a volume reduction and recovery process for 238 Pu contaminated waste. Combustible low-level waste material contaminated with 238 Pu residue is destroyed by oxidation in a 900 C molten salt reaction vessel. The combustible waste is destroyed creating carbon dioxide and steam and a small amount of ash and insoluble 2328 Pu in the spent salt. The valuable 238 Pu is recycled using aqueous recovery techniques. Experimental test results for this technology indicate a plutonium recovery efficiency of 99%. Molten salt oxidation stabilizes the waste converting it to a non-combustible waste. Thus installation and use of molten salt oxidation technology will substantially reduce the volume of 238 Pu contaminated waste. Cost-effectiveness evaluations of molten salt oxidation indicate a significant cost savings when compared to the present plans to package, or re-package, certify and transport these wastes to the Waste Isolation Pilot Plant for permanent disposal. Clear and distinct cost advantages exist for MSO when the monetary value of the recovered 238 Pu is considered

  20. Risk Assessment Study of Fluoride Salts: Probability-Impact Matrix of Renal and Hepatic Toxicity Markers.

    Science.gov (United States)

    Usuda, Kan; Ueno, Takaaki; Ito, Yuichi; Dote, Tomotaro; Yokoyama, Hirotaka; Kono, Koichi; Tamaki, Junko

    2016-09-01

    The present risk assessment study of fluoride salts was conducted by oral administration of three different doses of sodium and potassium fluorides (NaF, KF) and zinc fluoride tetrahydrate (ZnF2 •4H2O) to male Wistar rats. The rats were divided into control and nine experimental groups, to which oral injections of 0.5 mL distilled water and 0.5 mL of fluoride solutions, respectively, were given. The dosage of fluoride compounds was adjusted to contain 2.1 mg (low-dose group, LG), 4.3 mg (mid-dose group, MG), and 5.4 mg fluoride per 200 g rat body weight (high-dose group, HG) corresponding to 5, 10, and 12.5 % of LD50 values for NaF. The 24-h urine volume, N-acetyl-β-D-glucosaminidase (NAG) and creatinine clearance (Ccr) were measured as markers of possible acute renal impact. The levels of alanine aminotransferase (ALT) and aspartate aminotransferase (AST) were determined in serum samples as markers of acute hepatic impact. The levels of serum and urinary fluoride were determined to evaluate fluoride bioavailability. The results reveal that higher doses of NaF, KF, and ZnF2 induced renal damage as indicated by higher urinary NAG (p fluoride is a potential, dose-dependent risk factor of renal tubular damage.

  1. Renal excretion of water in men under hypokinesia and physical exercise with fluid and salt supplementation

    Science.gov (United States)

    Zorbas, Yan G.; Federenko, Youri F.; Togawa, Mitsui N.

    It has been suggested that under hypokinesia (reduced number of steps/day) and intensive physical exercise, the intensification of fluid excretion in men is apparently caused as a result of the inability of the body to retain optimum amounts of water. Thus, to evaluate this hypothesis, studies were performed with the use of fluid and sodium chloride (NaCl) supplements on 12 highly trained physically healthy male volunteers aged 19-24 years under 364 days of hypokinesis (HK) and a set of intensive physical exercises (PE). They were divided into two groups with 6 volunteers per group. The first group of subjects were submitted to HK and took daily fluid and salt supplements in very small doses and the second group of volunteers were subjected to intensive PE and fluid-salt supplements. For the simulation of the hypokinetic effect, both groups of subjects were kept under an average of 4000 steps/day. During the prehypokinetic period of 60 days and under the hypokinetic period of 364 days water consumed and eliminated in urine by the men, water content in blood, plasma volume, rate of glomerular filtration, renal blood flow, osmotic concentration of urine and blood were measured. Under HK, the rate of renal excretion of water increased considerably in both groups. The additional fluid and salt intake failed to normalize water balance adequately under HK and PE. It was concluded that negative water balance evidently resulted not from shortage of water in the diet but from the inability of the body to retain optimum amounts of fluid under HK and a set of intensive PEs.

  2. Alternative Electrochemical Salt Waste Forms, Summary of FY2010 Results

    International Nuclear Information System (INIS)

    Riley, Brian J.; Rieck, Bennett T.; Crum, Jarrod V.; Matyas, Josef; McCloy, John S.; Sundaram, S.K.; Vienna, John D.

    2010-01-01

    In FY2009, PNNL performed scoping studies to qualify two waste form candidates, tellurite (TeO2-based) glasses and halide minerals, for the electrochemical waste stream for further investigation. Both candidates showed promise with acceptable PCT release rates and effective incorporation of the 10% fission product waste stream. Both candidates received reprisal for FY2010 and were further investigated. At the beginning of FY2010, an in-depth literature review kicked off the tellurite glasses study. The review was aimed at ascertaining the state-of-the-art for chemical durability testing and mixed chloride incorporation for tellurite glasses. The literature review led the authors to 4 unique binary and 1 unique ternary systems for further investigation which include TeO2 plus the following: PbO, Al2O3-B2O3, WO3, P2O5, and ZnO. Each system was studied with and without a mixed chloride simulated electrochemical waste stream and the literature review provided the starting points for the baseline compositions as well as starting points for melting temperature, compatible crucible types, etc. The most promising glasses in each system were scaled up in production and were analyzed with the Product Consistency Test, a chemical durability test. Baseline and PCT glasses were analyzed to determine their state, i.e., amorphous, crystalline, phase separated, had undissolved material within the bulk, etc. Conclusions were made as well as the proposed direction for FY2011 plans. Sodalite was successfully synthesized by the sol-gel method. The vast majority of the dried sol-gel consisted of sodalite with small amounts of alumino-silicates and unreacted salt. Upon firing the powders made by sol-gel, the primary phase observed was sodalite with the addition of varying amounts of nepheline, carnegieite, lithium silicate, and lanthanide oxide. The amount of sodalite, nepheline, and carnegieite as well as the bulk density of the fired pellets varied with firing temperature, sol

  3. Alternative Electrochemical Salt Waste Forms, Summary of FY2010 Results

    Energy Technology Data Exchange (ETDEWEB)

    Riley, Brian J.; Rieck, Bennett T.; Crum, Jarrod V.; Matyas, Josef; McCloy, John S.; Sundaram, S. K.; Vienna, John D.

    2010-08-01

    In FY2009, PNNL performed scoping studies to qualify two waste form candidates, tellurite (TeO2-based) glasses and halide minerals, for the electrochemical waste stream for further investigation. Both candidates showed promise with acceptable PCT release rates and effective incorporation of the 10% fission product waste stream. Both candidates received reprisal for FY2010 and were further investigated. At the beginning of FY2010, an in-depth literature review kicked off the tellurite glasses study. The review was aimed at ascertaining the state-of-the-art for chemical durability testing and mixed chloride incorporation for tellurite glasses. The literature review led the authors to 4 unique binary and 1 unique ternary systems for further investigation which include TeO2 plus the following: PbO, Al2O3-B2O3, WO3, P2O5, and ZnO. Each system was studied with and without a mixed chloride simulated electrochemical waste stream and the literature review provided the starting points for the baseline compositions as well as starting points for melting temperature, compatible crucible types, etc. The most promising glasses in each system were scaled up in production and were analyzed with the Product Consistency Test, a chemical durability test. Baseline and PCT glasses were analyzed to determine their state, i.e., amorphous, crystalline, phase separated, had undissolved material within the bulk, etc. Conclusions were made as well as the proposed direction for FY2011 plans. Sodalite was successfully synthesized by the sol-gel method. The vast majority of the dried sol-gel consisted of sodalite with small amounts of alumino-silicates and unreacted salt. Upon firing the powders made by sol-gel, the primary phase observed was sodalite with the addition of varying amounts of nepheline, carnegieite, lithium silicate, and lanthanide oxide. The amount of sodalite, nepheline, and carnegieite as well as the bulk density of the fired pellets varied with firing temperature, sol

  4. Criticality considerations for salt-cake disolution in DOE waste tanks

    International Nuclear Information System (INIS)

    Trumble, E.F.; Niemer, K.A.

    1995-01-01

    A large amount of high-level waste is being stored in the form of salt cake at the Savannah River site (SRS) in large (1.3 x 106 gal) underground tanks awaiting startup of the Defense Waste Processing Facility (DWPF). This salt cake will be dissolved with water, and the solution will be fed to DWPF for immobilization in borosilicate glass. Some of the waste that was transferred to the tanks contained enriched uranium and plutonium from chemical reprocessing streams. As water is added to these tanks to dissolve the salt cake, the insoluble portion of this fissile material will be left behind in the tank as the salt solution is pumped out. Because the salt acts as a diluent to the fissile material, the process of repeated water addition, salt dissolution, and salt solution removal will act as a concentrating mechanism for the undissolved fissile material that will remain in the tank. It is estimated that tank 41 H at SRS contains 20 to 120 kg of enriched uranium, varying from 10 to 70% 235 U, distributed nonuniformly throughout the tank. This paper discusses the criticality concerns associated with the dissolution of salt cake in this tank. These concerns are also applicable to other salt cake waste tanks that contain significant quantities of enriched uranium and/or plutonium

  5. Treatment of waste salt from the advanced spent fuel conditioning process (II) : optimum immobilization condition

    International Nuclear Information System (INIS)

    Kim, Jeong Guk; Lee, Jae Hee; Yoo, Jae Hyung; Kim, Joon Hyung

    2004-01-01

    Since zeolite is known to be stable at a high temperature, it has been reported as a promising immobilization matrix for waste salt. The crystal structure of dehydrated zeolite A breaks down above 1060 K, resulting in the formation of an amorphous solid and re-crystallization to beta-Cristobalite. This structural degradation depends on the existence of chlorides. When contacted to HCl, zeolite 4A is not stable even at 473 K. The optimum consolidation condition for LiCl salt waste from the oxide fuel reduction process based on the electrochemical method (Advanced spent fuel Conditioning Process; ACP) has been studied using zeolite A since 2001. Actually the constituents of waste salt are water-soluble. And, alkali halides are known to be readily radiolyzed to yield interstitial halogens and metal colloids. For disposal in a geological repository, the waste salt must meet the acceptance criteria. For a waste form containing chloride salt, two of the more important criteria are leach resistance and waste form durability. In this work, we prepared some samples with different mixing ratios of LiCl salt to zeolite A, and then compared some characteristics such as thermal stability, salt occlusion, free chloride content, leach resistance, mixing effect, etc

  6. Reconsolidation of salt as applied to permanent seals for the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    Hansen, F.D.; Callahan, G.D.; Van Sembeek, L.L.

    1993-01-01

    Reconsolidated salt is a fundamental component of the permanent seals for the Waste Isolation Pilot Plant. As regulations are currently understood and seal concepts envisioned, emplaced salt is the sole long-term seal component designed to prevent the shafts from becoming preferred pathways for rating gases or liquids. Studies under way in support of the sealing function of emplaced salt include laboratory testing of crushed salt small-scale in situ tests, constitutive modeling of crushed salt, calculations of the opening responses during operation and closure, and design practicalities including emplacement techniques. This paper briefly summarizes aspects of these efforts and key areas of future work

  7. Saltstone: cement-based waste form for disposal of Savannah River Plant low-level radioactive salt waste

    International Nuclear Information System (INIS)

    Langton, C.A.

    1984-01-01

    Defense waste processing at the Savannah River Plant will include decontamination and disposal of approximately 400 million liters of waste containing NaNO 3 , NaOH, Na 2 SO 4 , and NaNO 2 . After decontamination, the salt solution is classified as low-level waste. A cement-based waste form, saltstone, has been designed for disposal of Savannah River Plant low-level radioactive salt waste. Bulk properties of this material have been tailored with respect to salt leach rate, permeability, and compressive strength. Microstructure and mineralogy of leached and unleached specimens were characterized by SEM and x-ray diffraction analyses. The disposal system for the DWPF salt waste includes reconstitution of the crystallized salt as a solution containing 32 wt % solids. This solution will be decontaminated to remove 137 Cs and 90 Sr and then stabilized in a cement-based waste form. Laboratory and field tests indicate that this stabilization process greatly reduces the mobility of all of the waste constitutents in the surface and near-surface environment. Engineered trenches for subsurface burial of the saltstone have been designed to ensure compatibility between the waste form and the environment. The total disposal sytem, saltstone-trench-surrounding soil, has been designed to contain radionuclides, Cr, and Hg by both physical encapsulation and chemical fixation mechanisms. Physical encapsulation of the salts is the mechanism employed for controlling N and OH releases. In this way, final disposal of the SRP low-level waste can be achieved and the quality of the groundwater at the perimeter of the disposal site meets EPA drinking water standards

  8. Assessment of lead tellurite glass for immobilizing electrochemical salt wastes from used nuclear fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Riley, Brian J.; Kroll, Jared O.; Peterson, Jacob A.; Pierce, David A.; Ebert, William L.; Williams, Benjamin D.; Snyder, Michelle M. V.; Frank, Steven M.; George, Jaime L.; Kruska, Karen

    2017-11-01

    This paper provides an overview of research evaluating the use of lead tellurite glass as a waste form for salt wastes from electrochemical reprocessing of used nuclear fuel. The efficacy of using lead tellurite glass to immobilize three different salt compositions was evaluated: a LiCl-Li2O oxide reduction salt containing fission products from oxide fuel, a LiCl-KCl eutectic salt containing fission products from metallic fuel, and SrCl2. Physical and chemical properties of glasses made with these salts were characterized with X-ray diffraction, bulk density measurements, differential thermal analysis, chemical durability tests, scanning and transmission electron microscopies, and energy-dispersive X-ray spectroscopy. These glasses were found to accommodate high salt concentrations and have high densities, but further development is needed to improve chemical durability. (C) 2017 Published by Elsevier B.V.

  9. Salt splitting of sodium-dominated radioactive waste using ceramic membranes

    International Nuclear Information System (INIS)

    Hollenberg, G.W.; Carlson, C.D.; Virkar, A.; Joshi, A.

    1994-08-01

    The potential for salt splitting of sodium dominated radioactive wastes by use of a ceramic membrane is reviewed. The technical basis for considering this processing technology is derived from the technology developed for battery and chlor-alkali chemical industry. Specific comparisons are made with the commercial organic membranes which are the standard in nonradioactive salt splitting. Two features of ceramic membranes are expected to be especially attractive: high tolerance to gamma irradiation and high selectivity between sodium and other ions. The objective of the salt splitting process is to separate nonradioactive sodium from contaminated sodium salts prior to other pretreatment processes in order to: (1) concentrate the waste in order to reduce the volume of subsequent additives and capacity of equipment, (2) decrease the pH of the waste in preparation for further processing, and (3) provide sodium with very low radioactivity levels for caustic washing of sludge or low level and mixed waste vitrification

  10. The thermo-mechanical behaviour of a salt dome with a heat-generating waste repository

    International Nuclear Information System (INIS)

    Janssen, L.G.J.; Prij, J.; Kevenaar, J.W.A.M.; Jong, C.J.T.; Klok, J.; Beemsterboer, C.

    1984-01-01

    This report reviews the analytical work on the disposal of radioactive waste in salt domes performed at ECN in the period 1 January 1980 to 31 December 1982. Chapter 4 in the main report covers the global temperature and deformation analyses of the salt dome and the surrounding rocks. The attached three topical reports cover self-contained parts of the study. The computer program TASTE developed to analyse, at acceptable cost and with, for engineering purposes, sufficient accuracies, the temperature rises in the salt dome due to the stored heat-generating waste is described in Annex 1. Annex 2 gives a description of the extended finite element program GOLIA. The program has been extended to make it suitable for the creep analysis of salt domes with repositories of heat-generating waste. The study on the closing and sealing of boreholes wit heat-generating waste is reported in Annex 3

  11. Analysis of the geological stability of a hypothetical radioactive waste repository in a bedded salt formation

    International Nuclear Information System (INIS)

    Tierney, M.S.; Lusso, F.; Shaw, H.R.

    1978-01-01

    This document reports on the development of mathematical models used in preliminary studies of the long-term safety of radioactive wastes deeply buried in bedded salt formations. Two analytical approaches to estimating the geological stability of a waste repository in bedded salt are described: (a) use of probabilistic models to estimate the a priori likelihoods of release of radionuclides from the repository through certain idealized natural and anthropogenic causes, and (b) a numerical simulation of certain feedback effects of emplacement of waste materials upon ground-water access to the repository's host rocks. These models are applied to an idealized waste repository for the sake of illustration

  12. Analyses of SRS waste glass buried in granite in Sweden and salt in the United States

    International Nuclear Information System (INIS)

    Williams, J.P.; Wicks, G.G.; Clark, D.E.; Lodding, A.R.

    1991-01-01

    Simulated Savannah River Site (SRS) waste glass forms have been buried in the granite geology of the Stirpa mine in Sweden for two years. Analyses of glass surfaces provided a measure of the performance of the waste glasses as a function of time. Similar SRS waste glass compositions have also been buried in salt at the WIPP facility in Carlsbad, New Mexico for a similar time period. Analyses of the SRS waste glasses buried in-situ in granite will be presented and compared to the performance of these same compositions buried in salt at WIPP

  13. Modeling of waste/near field interactions for a waste repository in bedded salt: the Dynamic Network (DNET) model

    International Nuclear Information System (INIS)

    Cranwell, R.M.

    1983-01-01

    The Fuel Cycle Risk Analysis Division of Sandia National Laboratories has been funded by the US Nuclear Regulatory Commission to develop a methodology for use in assessing the long-term risk from the disposal of radioactive wastes in deep geologic formations. As part of this program, the Dynamic Network (DNET) model was developed to investigate waste/near field interactions associated with the disposal of radioactive wastes in bedded salt formations. The model is a quasi-multi-dimensional network model with capabilities for simulating processes such as fluid flow, heat transport, salt dissolution, salt creep, and the effects of thermal expansion and subsedence on the rock units surrounding the repository. The use of DNET has been demonstrated in the analysis of a hypothetical disposal site containing a bedded salt formation as the host medium for the repository. An example of this demonstration analysis is discussed. Furthermore, the outcome of sensitivity analyses performed on the DNET model are presented

  14. Volume reduction of waste contaminated by fission product elements and plutonium using molten salt combustion

    International Nuclear Information System (INIS)

    McKenzie, D.E.; Grantham, L.F.; Paulson, R.B.

    1979-01-01

    In the Molten Salt Combustion Process, transuranic or β-γ organic waste and air are continuously introduced beneath the surface of a sodium carbonate-containing melt at a temperature of about 800 0 C. Complete combustion of the organic material to carbon dioxide and steam occurs without the conversion of nitrogen to nitrogen oxides. The noxious gases formed by combustion of the chloride, sulfur or phosphorus content of the waste instantly react with the melt to form the corresponding sodium compounds. These compounds as well as the ash and radionuclides are retained in the molten salt. The spent salt is either fused cast into an engineered disposal container or processed to recover salt and plutonium. Molten salt combustion reduces the waste to about 2% of its original volume. Many reactor or reprocessing wastes which cannot be incinerated without difficulty are readily combusted in the molten salt. A 50 kg/hr molten salt combustion system is being designed for the Radioactive Waste Management Complex at the Idaho National Engineering Laboratory. Construction of the combustor started during 1977, and combustor startup was scheduled for the spring of 1978

  15. Application of molten salt oxidation for the minimization and recovery of plutonium-238 contaminated wastes

    Energy Technology Data Exchange (ETDEWEB)

    Wishau, R.

    1998-05-01

    Molten salt oxidation (MSO) is proposed as a {sup 238}Pu waste treatment technology that should be developed for volume reduction and recovery of {sup 238}Pu and as an alternative to the transport and permanent disposal of {sup 238}Pu waste to the WIPP repository. In MSO technology, molten sodium carbonate salt at 800--900 C in a reaction vessel acts as a reaction media for wastes. The waste material is destroyed when injected into the molten salt, creating harmless carbon dioxide and steam and a small amount of ash in the spent salt. The spent salt can be treated using aqueous separation methods to reuse the salt and to recover 99.9% of the precious {sup 238}Pu that was in the waste. Tests of MSO technology have shown that the volume of combustible TRU waste can be reduced by a factor of at least twenty. Using this factor the present inventory of 574 TRU drums of {sup 238}Pu contaminated wastes is reduced to 30 drums. Further {sup 238}Pu waste costs of $22 million are avoided from not having to repackage 312 of the 574 drums to a drum total of more than 4,600 drums. MSO combined with aqueous processing of salts will recover approximately 1.7 kilograms of precious {sup 238}Pu valued at 4 million dollars (at $2,500/gram). Thus, installation and use of MSO technology at LANL will result in significant cost savings compared to present plans to transport and dispose {sup 238}Pu TRU waste to the WIPP site. Using a total net present value cost for the MSO project as $4.09 million over a five-year lifetime, the project can pay for itself after either recovery of 1.6 kg of Pu or through volume reduction of 818 drums or a combination of the two. These savings show a positive return on investment.

  16. Application of molten salt oxidation for the minimization and recovery of plutonium-238 contaminated wastes

    International Nuclear Information System (INIS)

    Wishau, R.

    1998-05-01

    Molten salt oxidation (MSO) is proposed as a 238 Pu waste treatment technology that should be developed for volume reduction and recovery of 238 Pu and as an alternative to the transport and permanent disposal of 238 Pu waste to the WIPP repository. In MSO technology, molten sodium carbonate salt at 800--900 C in a reaction vessel acts as a reaction media for wastes. The waste material is destroyed when injected into the molten salt, creating harmless carbon dioxide and steam and a small amount of ash in the spent salt. The spent salt can be treated using aqueous separation methods to reuse the salt and to recover 99.9% of the precious 238 Pu that was in the waste. Tests of MSO technology have shown that the volume of combustible TRU waste can be reduced by a factor of at least twenty. Using this factor the present inventory of 574 TRU drums of 238 Pu contaminated wastes is reduced to 30 drums. Further 238 Pu waste costs of $22 million are avoided from not having to repackage 312 of the 574 drums to a drum total of more than 4,600 drums. MSO combined with aqueous processing of salts will recover approximately 1.7 kilograms of precious 238 Pu valued at 4 million dollars (at $2,500/gram). Thus, installation and use of MSO technology at LANL will result in significant cost savings compared to present plans to transport and dispose 238 Pu TRU waste to the WIPP site. Using a total net present value cost for the MSO project as $4.09 million over a five-year lifetime, the project can pay for itself after either recovery of 1.6 kg of Pu or through volume reduction of 818 drums or a combination of the two. These savings show a positive return on investment

  17. A familial disorder with low bone density and renal phosphate wasting.

    NARCIS (Netherlands)

    Grondel, I.M.; Deure, J. van der; Zanen, A.L.; Dogger, M.; Heuvel, L.P.W.J. van den

    2009-01-01

    Hereditary forms of renal phosphate wasting have been studied thoroughly in the past years. X-linked Hypophosphatemic rickets (XLH), autosomal dominant hypophosphatemic rickets/osteomalacia (ADHR) and autosomal recessive hypophosphatemic rickets (ARHR) are known genetic disorders in which a

  18. Destruction of high explosives and wastes containing high explosives using the molten salt destruction process

    International Nuclear Information System (INIS)

    Upadhye, R.S.; Brummond, W.A.; Pruneda, C.O.

    1992-01-01

    This paper reports the Molten Salt Destruction (MSD) Process which has been demonstrated for the destruction of HE and HE-containing wastes. MSD has been used by Rockwell International and by Anti-Pollution Systems to destroy hazardous wastes. MSD converts the organic constituents (including the HE) of the waste into non-hazardous substances such as carbon dioxide, nitrogen and water. In the case of HE-containing mixed wastes, any actinides in the waste are retained in the molten salt, thus converting the mixed wastes into low-level wastes. (Even though the MSD process is applicable to mixed wastes, this paper will emphasize HE-treatment.) The destruction of HE is accomplished by introducing it, together with oxidant gases, into a crucible containing a molten salt, such as sodium carbonate, or a suitable mixture of the carbonates of sodium, potassium, lithium and calcium. The temperature of the molten salt can be between 400 to 900 degrees C. The combustible organic components of the waste react with oxygen to produce carbon dioxide, nitrogen and steam

  19. Risk assessment of nonhazardous oil-field waste disposal in salt caverns

    International Nuclear Information System (INIS)

    Elcock, D.

    1998-01-01

    Salt caverns can be formed in underground salt formations incidentally as a result of mining or intentionally to create underground chambers for product storage or waste disposal. For more than 50 years, salt caverns have been used to store hydrocarbon products. Recently, concerns over the costs and environmental effects of land disposal and incineration have sparked interest in using salt caverns for waste disposal. Countries using or considering using salt caverns for waste disposal include Canada (oil-production wastes), Mexico (purged sulfates from salt evaporators), Germany (contaminated soils and ashes), the United Kingdom (organic residues), and the Netherlands (brine purification wastes). In the US, industry and the regulatory community are pursuing the use of salt caverns for disposal of oil-field wastes. In 1988, the US Environmental Protection Agency (EPA) issued a regulatory determination exempting wastes generated during oil and gas exploration and production (oil-field wastes) from federal hazardous waste regulations--even though such wastes may contain hazardous constituents. At the same time, EPA urged states to tighten their oil-field waste management regulations. The resulting restrictions have generated industry interest in the use of salt caverns for potentially economical and environmentally safe oil-field waste disposal. Before the practice can be implemented commercially, however, regulators need assurance that disposing of oil-field wastes in salt caverns is technically and legally feasible and that potential health effects associated with the practice are acceptable. In 1996, Argonne National Laboratory (ANL) conducted a preliminary technical and legal evaluation of disposing of nonhazardous oil-field wastes (NOW) into salt caverns. It investigated regulatory issues; the types of oil-field wastes suitable for cavern disposal; cavern design and location considerations; and disposal operations, closure and remediation issues. It determined

  20. Risk assessment of nonhazardous oil-field waste disposal in salt caverns.

    Energy Technology Data Exchange (ETDEWEB)

    Elcock, D.

    1998-03-10

    Salt caverns can be formed in underground salt formations incidentally as a result of mining or intentionally to create underground chambers for product storage or waste disposal. For more than 50 years, salt caverns have been used to store hydrocarbon products. Recently, concerns over the costs and environmental effects of land disposal and incineration have sparked interest in using salt caverns for waste disposal. Countries using or considering using salt caverns for waste disposal include Canada (oil-production wastes), Mexico (purged sulfates from salt evaporators), Germany (contaminated soils and ashes), the United Kingdom (organic residues), and the Netherlands (brine purification wastes). In the US, industry and the regulatory community are pursuing the use of salt caverns for disposal of oil-field wastes. In 1988, the US Environmental Protection Agency (EPA) issued a regulatory determination exempting wastes generated during oil and gas exploration and production (oil-field wastes) from federal hazardous waste regulations--even though such wastes may contain hazardous constituents. At the same time, EPA urged states to tighten their oil-field waste management regulations. The resulting restrictions have generated industry interest in the use of salt caverns for potentially economical and environmentally safe oil-field waste disposal. Before the practice can be implemented commercially, however, regulators need assurance that disposing of oil-field wastes in salt caverns is technically and legally feasible and that potential health effects associated with the practice are acceptable. In 1996, Argonne National Laboratory (ANL) conducted a preliminary technical and legal evaluation of disposing of nonhazardous oil-field wastes (NOW) into salt caverns. It investigated regulatory issues; the types of oil-field wastes suitable for cavern disposal; cavern design and location considerations; and disposal operations, closure and remediation issues. It determined

  1. Some geotechnical problems related to underground waste disposal in salt formations

    International Nuclear Information System (INIS)

    Berest, P.

    1993-01-01

    Nuclear waste disposal in deep salt formations is an option considered by several countries. Rock salt is a very impervious medium, but can be easily leached; selection of an appropriate disposal formation must account for natural protections of the formation as regards water movements. It must be checked that such initially favourable characteristics will not be affected by the existence of shafts and galleries, or by the important heat output generated by vitrified wastes. The discussion is uneasy, for a comprehensive rheological model for rock salt is difficult to set and to be extrapolated to large time scales; some methodological problems are raised by use of numerical computations. (author). 22 refs., 2 figs

  2. Brine migration in salt and its implications in the geologic disposal of nuclear waste

    International Nuclear Information System (INIS)

    Jenks, G.H.; Claiborne, H.C.

    1981-12-01

    This report respresents a comprehensive review and analysis of available information relating to brine migration in salt surrounding radioactive waste in a salt repository. The topics covered relate to (1) the characteristics of salt formations and waste packages pertinent to considerations of rates, amounts, and effects of brine migration, (2) experimental and theoretical information on brine migration, and (3) means of designing to minimize any adverse effects of brine migration. Flooding, brine pockets, and other topics were not considered, since these features will presumably be eliminated by appropriate site selection and repository design. 115 references

  3. High salt inclusion reduces concentrate intake without major effects on renal function in young bulls

    Directory of Open Access Journals (Sweden)

    Mireia Blanco

    2014-08-01

    Full Text Available Beef producers prefer to feed concentrates on an ad libitum basis to increase the flexibility of their work. Including salt, which is a self-limiting supplement, could control or reduce concentrate intake without increasing the workforce. The aim of the study was to evaluate the effect of including 10%NaCl in the concentrate on intake, growth, blood ions (sodium, potassium and chlorine, renal function (through creatinine and urea concentrations in blood, and daytime behaviour of bulls over 6 weeks. Bulls consuming the control concentrate (Control bulls had greater weight gain (P<0.05 and concentrate intake (P<0.001 than those consuming the concentrate with 10%NaCl (10%NaCl bulls. Lower plasma sodium concentration was found in Control bulls after 6 weeks (P<0.05, while potassium concentration was lower after 4 (P<0.05 and 6 weeks (P<0.01. Blood urea did not differ between the groups, and creatinine only differed at week 4 (P<0.01. Control bulls spent less time eating hay (P<0.001 and more time idling (P<0.01 during daylight hours. In conclusion, the inclusion of 10%NaCl in the concentrate for short periods could be used to reduce concentrate intake without major effects on renal function; however, a concomitant decrease in weight gain should be expected.

  4. Using Aspen simulation package to determine solubility of mixed salts in TRU waste evaporator bottoms

    Energy Technology Data Exchange (ETDEWEB)

    Hatchell, J.L.

    1998-03-01

    Nitric acid from plutonium process waste is a candidate for waste minimization by recycling. Process simulation software packages, such as Aspen, are valuable tools to estimate how effective recovery processes can be, however, constants in equations of state for many ionic components are not in their data libraries. One option is to combine single salt solubility`s in the Aspen model for mixed salt system. Single salt solubilities were regressed in Aspen within 0.82 weight percent of literature values. These were combined into a single Aspen model and used in the mixed salt studies. A simulated nitric acid waste containing mixed aluminum, calcium, iron, magnesium and sodium nitrate was tested to determine points of solubility between 25 and 100 C. Only four of the modeled experimental conditions, at 50 C and 75 C, produced a saturated solution. While experimental results indicate that sodium nitrate is the first salt to crystallize out, the Aspen computer model shows that the most insoluble salt, magnesium nitrate, the first salt to crystallize. Possible double salt formation is actually taking place under experimental conditions, which is not captured by the Aspen model.

  5. Nuclear waste repository simulation experiments, Asse Salt Mine, Federal Republic of Germany. Annual report, 1983

    International Nuclear Information System (INIS)

    Rothfuchs, T.; Luebker, D.; Coyle, A.; Kalia, H.

    1984-10-01

    This is the First Annual report (1983) which describes experiments simulating a nuclear waste respository at the 800-meter level of the Asse Salt Mine in the Federal Republic of Germany. The report describes the test equipment, the Asse Salt Mine, the pretest properties of the salt in the test gallery, and the mine proper. Also included are test data for the first six months of operations on brine migration rates, room closure rates, extensometer readings, stress measurements, and thermal mechanical behavior of the salt. The duration of the experiments will be two years, ending in December 1985. 3 references, 34 figures, 13 tables

  6. Alternative Electrochemical Salt Waste Forms, Summary of FY/CY2011 Results

    Energy Technology Data Exchange (ETDEWEB)

    Riley, Brian J.; McCloy, John S.; Crum, Jarrod V.; Rodriguez, Carmen P.; Windisch, Charles F.; Lepry, William C.; Matyas, Josef; Westman, Matthew P.; Rieck, Bennett T.; Lang, Jesse B.; Pierce, David A.

    2011-12-01

    This report summarizes the 2011 fiscal+calendar year efforts for developing waste forms for a spent salt generated in reprocessing nuclear fuel with an electrochemical separations process. The two waste forms are tellurite (TeO2-based) glasses and sol-gel-derived high-halide mineral analogs to stable minerals found in nature.

  7. Alternative Electrochemical Salt Waste Forms, Summary of FY/CY2011 Results

    International Nuclear Information System (INIS)

    Riley, Brian J.; McCloy, John S.; Crum, Jarrod V.; Rodriguez, Carmen P.; Windisch, Charles F.; Lepry, William C.; Matyas, Josef; Westman, Matthew P.; Rieck, Bennett T.; Lang, Jesse B.; Pierce, David A.

    2011-01-01

    This report summarizes the 2011 fiscal+calendar year efforts for developing waste forms for a spent salt generated in reprocessing nuclear fuel with an electrochemical separations process. The two waste forms are tellurite (TeO2-based) glasses and sol-gel-derived high-halide mineral analogs to stable minerals found in nature.

  8. Systems costs for disposal of Savannah River high-level waste sludge and salt

    International Nuclear Information System (INIS)

    McDonell, W.R.; Goodlett, C.B.

    1984-01-01

    A systems cost model has been developed to support disposal of defense high-level waste sludge and salt generated at the Savannah River Plant. Waste processing activities covered by the model include decontamination of the salt by a precipitation process in the waste storage tanks, incorporation of the sludge and radionuclides removed from the salt into glass in the Defense Waste Processing Facility (DWPF), and, after interim storage, final disposal of the DWPF glass waste canisters in a federal geologic repository. Total costs for processing of waste generated to the year 2000 are estimated to be about $2.9 billion (1984 dollars); incremental unit costs for DWPF and repository disposal activities range from $120,000 to $170,000 per canister depending on DWPF processing schedules. In a representative evaluation of process alternatives, the model is used to demonstrate cost effectiveness of adjustments in the frit content of the waste glass to reduce impacts of wastes generated by the salt decontamination operations. 13 references, 8 tables

  9. Assessment of lead tellurite glass for immobilizing electrochemical salt wastes from used nuclear fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Riley, Brian J.; Kroll, Jared O.; Peterson, Jacob A.; Pierce, David A.; Ebert, William L.; Williams, Benjamin D.; Snyder, Michelle M. V.; Frank, Steven M.; George, Jaime L.; Kruska, Karen

    2017-11-01

    This paper provides an overview of research evaluating the use of tellurite glass as a waste form for salt wastes from electrochemical processing. The capacities to immobilize different salts were evaluated including: a LiCl-Li2O oxide reduction salt (for oxide fuel) containing fission products, a LiCl-KCl eutectic salt (for metallic fuel) containing fission products, and SrCl2. Physical and chemical properties of the glasses were characterized by using X-ray diffraction, bulk density measurements, chemical durability tests, scanning electron microscopy, and energy dispersive X-ray emission spectroscopy. These glasses were found to accommodate high concentrations of halide salts and have high densities. However, improvements are needed to meet chemical durability requirements.

  10. Secondary Aluminum Processing Waste: Salt Cake Characterization and Reactivity

    Science.gov (United States)

    Thirty-nine salt cake samples were collected from 10 SAP facilities across the U.S. The facilities were identified by the Aluminum Association to cover a wide range of processes. Results suggest that while the percent metal leached from the salt cake was relatively low, the leac...

  11. Test procedures for polyester immobilized salt-containing surrogate mixed wastes

    International Nuclear Information System (INIS)

    Biyani, R.K.; Hendrickson, D.W.

    1997-01-01

    These test procedures are written to meet the procedural needs of the Test Plan for immobilization of salt containing surrogate mixed waste using polymer resins, HNF-SD-RE-TP-026 and to ensure adequacy of conduct and collection of samples and data. This testing will demonstrate the use of four different polyester vinyl ester resins in the solidification of surrogate liquid and dry wastes, similar to some mixed wastes generated by DOE operations

  12. Comparison of slagging pyrolysis and molten salt incinerators for treating TRU waste at the INEL

    International Nuclear Information System (INIS)

    1977-11-01

    For the comparison, it is assumed that the waste product is required to meet the acceptance criteria of the Waste Isolation Pilot Plant, i.e., low leachability. Slagging pyrolysis incinerates combustible waste and melts noncombustible waste; the resulting slag forms a glass of low leachability. In the molten salt incinerator, combustion occurs at low temperatures with no accumulation of explosive gases, but the waste must have been previously sorted into combustibles and noncombustibles and then shredded. The economics, safety, and technical features are compared. Advantages, disadvantages, and areas of technical uncertainty of the two systems are listed. Development costs and schedules for the two types of incinerators are discussed

  13. Solubility and speciation of actinides in salt solutions and migration experiments of intermediate level waste in salt formations

    International Nuclear Information System (INIS)

    1986-01-01

    A comprehensive study into the solubility of the actinides americium and plutonium in concentrated salt solutions, the release of radionuclides from various forms of conditioned ILW and the migration behaviour of these nuclides through geological material specific to the Gorleben site in Lower Saxony is described. A detailed investigation into the characterization of four highly concentrated salt solutions in terms of their pH, Eh, inorganic carbon contents and their densities is given and a series of experiments investigating the solubility of standard americium(III) and plutonium(IV) hydroxides in these solutions is described. Transuranic mobility studies for solutions derived from the standard hydroxides through salt and sand have shown the presence of at least two types of species present of widely differing mobility; one migrating with approximately the same velocity as the solvent front and the other strongly retarded. Actinide mobility data are presented and discussed for leachates derived from the simulated ILW in cement and data are also presented for the migration of the fission products in leachates derived from real waste solidified in cement and bitumen. Relatively high plutonium mobilities were observed in the case of the former and in the case of the real waste leachates, cesium was found to be the least retarded. The sorption of ruthenium was found to be largely associated with the insoluble residues of the natural rock salt rather than the halite itself. (orig./RB)

  14. Rock salt as a medium for long-term isolation of radioactive wastes - a reassessment

    International Nuclear Information System (INIS)

    Chaturvedi, L.

    1985-01-01

    Rock salt has been regarded as a suitable medium for the permanent disposal of high and medium level radioactive wastes since the National Academy of Sciences recommended it in 1957. As a result of detained site-specific studies conducted for the Waste Isolation Pilot Plant (WIPP) project in New Mexico, however, several potential problems which are unique to bedded salt deposits have emerged. These include 1) the need to delineate the extent and rate of past dissolution and projections for the future, 2) the origin and significance of brines often found underlying the salt beds, 3) the rate and volume of migration of brine from the salt crystals towards the heat producing waste canisters, 4) the creep rates and implications for retrievability, and 5) the existence of potash and oil and gas resources with implications of human intrusion in the future. These questions will also be faced for sites in salt domes with added complications due to more complex structure and hydrology. The experience at WIPP shows that the site characterization process for high level waste repositories in bedded or dome salt should aim at identifying the important issues of site suitability early in the process and a clear program should be established to address these issues

  15. Radiolysis salt phenomenology: application to storage of high level radioactive waste

    International Nuclear Information System (INIS)

    Akram, Najib

    1993-01-01

    In France, rock salt is a candidate repository for highly radioactive waste. Rock salt contains water and adsorbed gases which can be released in boreholes after heating due to vitrified wastes. In addition, waste-induced irradiation in near-field conditions induce radiolytic reactions which also contribute to gas release. The aim of this work is to understand and evaluate the effects of heat and irradiation produced by waste containers in a deep disposal, primarily concerning gas production. This is justified by the impact of gases on long-term safety: toxicity, explosibility, chemical reactivity, pressure build-up. We have evidenced the influence of integrated dose, filling gases, temperature and grain size on an homogeneous medium (Asse Mine rock salt). We have then studied heterogeneous samples, which allowed to determine the influence of the chemical and mineralogical composition of rock salt (bedded rock salt from the Mine de Potasse d'Alsace). The role played by organic matter on gas production is important, leading for instance to high consumption rates of oxygen. Through this study, we have also considered the behaviour of clay-rich materials under irradiation. Our results constitute important bases for the future modelling of the phenomena which will take place in the near-field of a rock salt-type repository, especially concerning its long-term safety. (author) [fr

  16. Radiological consequences associated with human intrusion into radioactive waste repositories in salt formations

    International Nuclear Information System (INIS)

    Jacquier, P.

    1989-01-01

    The assessment of the radiological impact of human intrusion scenarios is extremely important in the case of repositories located in salt formations, since salt is obviously a valuable economic resource. Salt formations also represent a suitable medium for mining storage caverns for oil and gas. The scenario considered in this report is that of solution mining in salt formations to produce salt for human consumption. It is postulated that the salt is extracted by excavating a cavern through solution-mining and that, in the course of cavern enlargement, the waste is intercepted and drops to the bottom of the cavern. We have assumed that the intrusion takes place 500 or even 2 500 years after the repository has been sealed. The cases considered involve high-level vitrified waste or cemented alpha waste. The paper describes the assumptions on which the scenario is based and uses a simplified model to assess the radiological consequences associated with the ingestion of contaminated salt. The paper also provides details of a sensitivity/uncertainty analysis which identified several areas in which experimental studies should be either initiated or continued [fr

  17. Radioactive waste isolation in salt: peer review of the Office of Nuclear Waste Isolation's report on Functional Design Criteria for a Repository for High-Level Radioactive Waste

    International Nuclear Information System (INIS)

    Hambley, D.F.; Russell, J.E.; Busch, J.S.; Harrison, W.; Edgar, D.E.; Tisue, M.W.

    1984-08-01

    This report summarizes Argonne's review of the Office of Nuclear Waste Isolation's (ONWI's) draft report entitled Functional Design Criteria for High-Level Nuclear Waste Repository in Salt, dated January 23, 1984. Recommendations are given for improving the ONWI draft report

  18. Study on application of molten salt oxidation technology (MSO) for PVC wastes treatment

    International Nuclear Information System (INIS)

    Tran Thu Ha; Nguyen Hong Quy; Pham Quoc Ky; Nguyen Quang Long; Vuong Thu Bac; Dang Duc Nhan

    2007-01-01

    The project 'Study on application of molten salt oxidation (MSO) for PVC plastic wastes treatment' aims at three followings: 1) Installation of lab-scale MSO unit with essential compositions builds up foundation for the 2) estimation of waste destruction efficiency of the technology. 3) Based on the results of testing PVC - the chlorinated organic wastes on the lab-scale unit, the ability of the technology application at pilot-scale level will be primary estimated. The adjustment and correction of some compositions in the lab-scale unit theoretically designed during experiment overcame the shortages by design and fabrication such as heat distribution regime, feeding wastes and draining spent salt. These solutions adapt to the technical requirement of operation as well as scientific requirement of the research on MSO process. PVC waste treatment was tested on the MSO lab-scale unit in different conditions of operation temperature, superficial air velocity related to air/oxygen feeding rate, waste feeding rate. The testing results showed that destruction efficiency of chlorine in MSO technology was almost absolute. HCl and Cl 2 emission were insignificant in different operation conditions. HCl and Cl 2 emission depend on resident time and nature of molten salt. However, with inherent attributes of MSO technology emission of CO is not avoided in processing waste treatment. Therefore, finding active solutions for reduction CO emission is essential to complete the technology. The experiments also were carried in conditions of single molten salt (Na 2 CO 3 ) and molten (Na 2 CO 3 - K 2 CO 3 ) eutectic. The comparison of efficiency of these tests gives idea of using molten salt eutectic to reduce operation cost in MSO technology. Based on operation parameters and scientific verification results during experiments, the introductory procedure of waste treatment by MSO process was built up. Thereby, primary estimation of development of the technology in pilot-scale is given

  19. Conceptual waste package interim product specifications and data requirements for disposal of borosilicate glass defense high-level waste forms in salt geologic repositories

    International Nuclear Information System (INIS)

    1983-06-01

    The conceptual waste package interim product specifications and data requirements presented are applicable specifically to the normal borosilicate glass product of the Defense Waste Processing Facility (DWPF). They provide preliminary numerical values for the defense high-level waste form parameters and properties identified in the waste form performance specification for geologic isolation in salt repositories. Subject areas treated include containment and isolation, operational period safety, criticality control, waste form/production canister identification, and waste package performance testing requirements. This document was generated for use in the development of conceptual waste package designs in salt. It will be revised as additional data, analyses, and regulatory requirements become available

  20. Modeling internal deformation of salt structures targeted for radioactive waste disposal

    International Nuclear Information System (INIS)

    Chemia, Zurab

    2008-01-01

    This thesis uses results of systematic numerical models to argue that externally inactive salt structures, which are potential targets for radioactive waste disposal, might be internally active due to the presence of dense layers or blocks within a salt layer. The three papers that support this thesis use the Gorleben salt diapir (NW Germany), which was targeted as a future final repository for high-grade radioactive waste, as a general guideline. The first two papers present systematic studies of the parameters that control the development of a salt diapir and how it entrains a dense anhydrite layer. Results from these numerical models show that the entrainment of a dense anhydrite layer within a salt diapir depends on four parameters: sedimentation rate, viscosity of salt, perturbation width and the stratigraphic location of the dense layer. The combined effect of these four parameters, which has a direct impact on the rate of salt supply (volume/area of the salt that is supplied to the diapir with time), shape a diapir and the mode of entrainment. Salt diapirs down-built with sedimentary units of high viscosity can potentially grow with an embedded anhydrite layer and deplete their source layer (salt supply ceases). However, when salt supply decreases dramatically or ceases entirely, the entrained anhydrite layer/segments start to sink within the diapir. In inactive diapirs, sinking of the entrained anhydrite layer is inevitable and strongly depends on the rheology of the salt, which is in direct contact with the anhydrite layer. During the post-depositional stage, if the effective viscosity of salt falls below the threshold value of around 10 18 -10 19 Pa s, the mobility of anhydrite blocks might influence any repository within the diapir. However, the internal deformation of the salt diapir by the descending blocks decreases with increase in effective viscosity of salt. The results presented in this thesis suggest that it is highly likely that salt structures

  1. Radiological consequences of a human intrusion in a nuclear waste repository in a salt formation

    International Nuclear Information System (INIS)

    Jacquier, P.; Raimbault, P.

    1989-07-01

    The assessment of the consequences of human intrusion scenarios for a repository is very important for salt formations, since this material has an undeniable economic interest. In this work, the scenario considers the solution mining of salt for human consumption: salt is extracted from a cavern; by leaching, this cavern enlarges and uncovers the waste, which falls down into the sump. It was assumed that the intrusion takes place either 500 years or 2500 years after the closing of the repository. High-level vitrified waste or alpha cemented waste were considered. This paper displays the assumptions made and, using a simplified modelling of the phenomena, the estimation of the radiological consequences due to ingestion of contamined sals. A sensitivity/uncertainty analysis is presented which emphasizes several fields where experimental studies have to be pursued or launched [fr

  2. Thermal decomposition of nitrate salts liquid waste for the lagoon sludge treatment

    International Nuclear Information System (INIS)

    Hwang, D. S.; Oh, J. H.; Kim, Y. K.; Lee, K. Y.; Choi, Y. D.; Hwang, S. T.; Park, J. H.

    2004-01-01

    This study investigated the thermal decomposition property of nitrate salts liquid waste which is produced in a series of the processes for the sludge treatment. Thermal decomposition property was analyzed by TG/DTA and XRD. Most ammonium nitrate in the nitrate salts liquid waste was decomposed at 250 .deg. C and calcium nitrate was decomposed and converted into calcium oxide at 550 .deg. C. Sodium nitrate was decomposed at 700 .deg. C and converted into sodium oxide which reacts with water easily. But sodium oxide was able to convert into a stable compound by adding alumina. Therefore, nitrate salts liquid waste can be treated by two steps as follows. First, ammonium nitrate is decomposed at 250 .deg. C. Second, alumina is added in residual solid sodium nitrate and calcium nitrate and these are decomposed at 900 .deg. C. Final residue consists of calcium oxide and Na 2 O.Al 2 O 3 and can be stored stably

  3. Salt Repository Project: Waste Package Program (WPP) modeling activiteis: FY 1984 annual report

    International Nuclear Information System (INIS)

    Kuhn, W.L.; Simonson, S.A.; Pulsipher, B.A.

    1987-03-01

    The Pacific Northwest Laboratory (PNL) is supporting the US Department of Energy's (DOE) Salt Repository Project (SRP) through its Waste Package Program (WPP). During FY 1984, the WPP continued its program of waste package component development and interactions testing and application of the resulting data base to develop predictive models describing waste package degradation and radionuclide release. Within the WPP, the Modeling Task (Task 04 during FY 1984) was conducted to interpret the tests in such a way that scientifically defensible models can be developed for use in qualification of the waste package

  4. Glovebox design requirements for molten salt oxidation processing of transuranic waste

    Energy Technology Data Exchange (ETDEWEB)

    Ramsey, K.B.; Acosta, S.V. [Los Alamos National Lab., NM (United States); Wernly, K.D. [Molten Salt Oxidation Corp., Bensalem, PA (United States)

    1998-12-31

    This paper presents an overview of potential technologies for stabilization of {sup 238}Pu-contaminated combustible waste. Molten salt oxidation (MSO) provides a method for removing greater than 99.999% of the organic matrix from combustible waste. Implementation of MSO processing at the Los Alamos National Laboratory (LANL) Plutonium Facility will eliminate the combustible matrix from {sup 238}Pu-contaminated waste and consequently reduce the cost of TRU waste disposal operations at LANL. The glovebox design requirements for unit operations including size reduction and MSO processing will be presented.

  5. Glovebox design requirements for molten salt oxidation processing of transuranic waste

    International Nuclear Information System (INIS)

    Ramsey, K.B.; Acosta, S.V.; Wernly, K.D.

    1998-01-01

    This paper presents an overview of potential technologies for stabilization of 238 Pu-contaminated combustible waste. Molten salt oxidation (MSO) provides a method for removing greater than 99.999% of the organic matrix from combustible waste. Implementation of MSO processing at the Los Alamos National Laboratory (LANL) Plutonium Facility will eliminate the combustible matrix from 238 Pu-contaminated waste and consequently reduce the cost of TRU waste disposal operations at LANL. The glovebox design requirements for unit operations including size reduction and MSO processing will be presented

  6. Congenital primary adrenal insufficiency and selective aldosterone defects presenting as salt-wasting in infancy: a single center 10-year experience.

    Science.gov (United States)

    Bizzarri, Carla; Olivini, Nicole; Pedicelli, Stefania; Marini, Romana; Giannone, Germana; Cambiaso, Paola; Cappa, Marco

    2016-08-02

    Salt-wasting represents a relatively common cause of emergency admission in infants and may result in life-threatening complications. Neonatal kidneys show low glomerular filtration rate and immaturity of the distal nephron leading to reduced ability to concentrate urine. A retrospective chart review was conducted for infants hospitalized in a single Institution from 1(st) January 2006 to 31(st) December 2015. The selection criterion was represented by the referral to the Endocrinology Unit for hyponatremia (serum sodium <130 mEq/L) of suspected endocrine origin at admission. Fifty-one infants were identified. In nine infants (17.6 %) hyponatremia was related to unrecognized chronic gastrointestinal or renal salt losses or reduced sodium intake. In 10 infants (19.6 %) hyponatremia was related to central nervous system diseases. In 19 patients (37.3 %) the final diagnosis was congenital adrenal hyperplasia (CAH). CAH was related to 21-hydroxylase deficiency in 18 patients, and to 3β-Hydroxysteroid dehydrogenase (3βHSD) deficiency in one patient. Thirteen patients (25.5 %) were affected by different non-CAH salt-wasting forms of adrenal origin. Four familial cases of X-linked adrenal hypoplasia congenita due to NROB1 gene mutation were identified. Two unrelated girls showed aldosterone synthase deficiency due to mutation of the CYP11B2 gene. Two unrelated infants were affected by familial glucocorticoid deficiency due to MC2R gene mutations. One girl showed pseudohypoaldosteronism related to mutations of the SCNN1G gene encoding for the epithelial sodium channel. Transient pseudohypoaldosteronism was identified in two patients with renal malformations. In two infants the genetic aetiology was not identified. Emergency management of infants presenting with salt wasting requires correction of water losses and treatment of electrolyte imbalances. Nevertheless, the differential diagnosis may be difficult in emergency settings, and sometimes hospitalized infants

  7. Long-term interactions of full-scale cemented waste simulates with salt brines

    Energy Technology Data Exchange (ETDEWEB)

    Kienzler, B.; Borkel, C.; Metz, V.; Schlieker, M.

    2016-07-01

    Since 1967 radioactive wastes have been disposed of in the Asse II salt mine in Northern Germany. A significant part of these wastes originated from the pilot reprocessing plant WAK in Karlsruhe and consisted of cemented NaNO{sub 3} solutions bearing fission products, actinides, as well as process chemicals. With respect to the long-term behavior of these wastes, the licensing authorities requested leaching experiments with full scale samples in relevant salt solutions which were performed since 1979. The experiments aimed at demonstrating the transferability of results obtained with laboratory samples to real waste forms and at the investigation of the effects of the industrial cementation process on the properties of the waste forms. This research program lasted until 2013. The corroding salt solutions were sampled several times and after termination of the experiments, the solid materials were analyzed by various methods. The results presented in this report cover the evolution of the solutions and the chemical and mineralogical characterization of the solids including radionuclides and waste components, and the paragenesis of solid phases (corrosion products). The outcome is compared to the results of model calculations. For safety analysis, conclusions are drawn on radionuclide retention, evolution of the geochemical environment, evolution of the density of solutions, and effects of temperature and porosity of the cement waste simulates on cesium mobilization.

  8. Long-term interactions of full-scale cemented waste simulates with salt brines

    International Nuclear Information System (INIS)

    Kienzler, B.; Borkel, C.; Metz, V.; Schlieker, M.

    2016-01-01

    Since 1967 radioactive wastes have been disposed of in the Asse II salt mine in Northern Germany. A significant part of these wastes originated from the pilot reprocessing plant WAK in Karlsruhe and consisted of cemented NaNO 3 solutions bearing fission products, actinides, as well as process chemicals. With respect to the long-term behavior of these wastes, the licensing authorities requested leaching experiments with full scale samples in relevant salt solutions which were performed since 1979. The experiments aimed at demonstrating the transferability of results obtained with laboratory samples to real waste forms and at the investigation of the effects of the industrial cementation process on the properties of the waste forms. This research program lasted until 2013. The corroding salt solutions were sampled several times and after termination of the experiments, the solid materials were analyzed by various methods. The results presented in this report cover the evolution of the solutions and the chemical and mineralogical characterization of the solids including radionuclides and waste components, and the paragenesis of solid phases (corrosion products). The outcome is compared to the results of model calculations. For safety analysis, conclusions are drawn on radionuclide retention, evolution of the geochemical environment, evolution of the density of solutions, and effects of temperature and porosity of the cement waste simulates on cesium mobilization.

  9. Backfill barriers for nuclear waste repositories in salt

    Energy Technology Data Exchange (ETDEWEB)

    Nowak, E J; Odoj, R; Merz, E [eds.

    1981-06-01

    Backfill mixtures surrounding the waste form and canister can provide a neutral or slightly acidic, potentially reducing environment, prevent convective aqueous flow, and act as an effective radionuclide migration barrier. Bentonite is likely to remain hydrothermally stable but potentially sensitive to waste package interactions which could alter the pH, the ratio of dissolved wires, or the sorption properties of radionuclide species.

  10. An improvement study on the closed chamber distillation system for recovery of renewable salts from salt wastes containing radioactive rare earth compounds

    International Nuclear Information System (INIS)

    Eun, H.C.; Cho, Y.Z.; Lee, T.K.; Kim, I.T.; Park, G.I.; Lee, H.S.

    2013-01-01

    In this paper, an improvement study on the closed chamber distillation system for recovery of renewable salts from salt wastes containing radioactive rare earth compounds was performed to determine optimum operating conditions. It was very important to maintain the pressure in the distillation chamber below 10 Torr for a high efficiency (salt recovery >99 %) of the salt distillation. This required increasing the salt vaporization and condensation rates in the distillation system. It was confirmed that vaporization and condensation rates could be improved controlling the given temperature of top of the condensation chamber. In the distillation tests of the salt wastes containing rare earth compounds, the operation time at a given temperature was greatly reduced changing the given temperature of top of the condensation chamber from 780 to 700 deg C. (author)

  11. Treatment Study Plan for Nitrate Salt Waste Remediation Revision 1.0

    Energy Technology Data Exchange (ETDEWEB)

    Juarez, Catherine L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Funk, David John [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Vigil-Holterman, Luciana R. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Naranjo, Felicia Danielle [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-03-07

    The two stabilization treatment methods that are to be examined for their effectiveness in the treatment of both the unremediated and remediated nitrate salt wastes include (1) the addition of zeolite and (2) cementation. Zeolite addition is proposed based on the results of several studies and analyses that specifically examined the effectiveness of this process for deactivating nitrate salts. Cementation is also being assessed because of its prevalence as an immobilization method used for similar wastes at numerous facilities around the DOE complex, including at Los Alamos. The results of this Treatment Study Plan will be used to provide the basis for a Resource Conservation and Recovery Act (RCRA) permit modification request of the LANL Hazardous Waste Facility Permit for approval by the New Mexico Environment Department-Hazardous Waste Bureau (NMED-HWB) of the proposed treatment process and the associated facilities.

  12. Protein removal from waste brines generated during ham salting through acidification and centrifugation.

    Science.gov (United States)

    Gutiérrez-Martínez, Maria del Rosario; Muñoz-Guerrero, Hernán; Alcaína-Miranda, Maria Isabel; Barat, José Manuel

    2014-03-01

    The salting step in food processes implies the production of large quantities of waste brines, having high organic load, high conductivity, and other pollutants with high oxygen demand. Direct disposal of the residual brine implies salinization of soil and eutrophication of water. Since most of the organic load of the waste brines comes from proteins leaked from the salted product, precipitation of dissolved proteins by acidification and removal by centrifugation is an operation to be used in waste brine cleaning. The aim of this study is optimizing the conditions for carrying out the separation of proteins from waste brines generated in the pork ham salting operation, by studying the influence of pH, centrifugal force, and centrifugation time. Models for determining the removal of proteins depending on the pH, centrifugal force, and time were obtained. The results showed a high efficacy of the proposed treatment for removing proteins, suggesting that this method could be used for waste brine protein removal. The best pH value to be used in an industrial process seems to be 3, while the obtained results indicate that almost 90% of the proteins from the brine can be removed by acidification followed by centrifugation. A further protein removal from the brine should have to be achieved using filtrating techniques, which efficiency could be highly improved as a consequence of the previous treatment through acidification and centrifugation. Waste brines from meat salting have high organic load and electrical conductivity. Proteins can be removed from the waste brine by acidification and centrifugation. The total protein removal can be up to 90% of the initial content of the waste brine. Protein removal is highly dependent on pH, centrifugation rate, and time. © 2014 Institute of Food Technologists®

  13. Alternative Electrochemical Salt Waste Forms, Summary of FY11-FY12 Results

    Energy Technology Data Exchange (ETDEWEB)

    Riley, Brian J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Mccloy, John S. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Crum, Jarrod V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lepry, William C. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Rodriguez, Carmen P. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Windisch, Charles F. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Matyas, Josef [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Westman, Matthew P. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Rieck, Bennett T. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lang, Jesse B. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Olszta, Matthew J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pierce, David A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2014-01-17

    The Fuel Cycle Research and Development Program, sponsored by the U.S. Department of Energy Office of Nuclear Energy, is currently investigating alternative waste forms for wastes generated from nuclear fuel processing. One such waste results from an electrochemical separations process, called the “Echem” process. The Echem process utilizes a molten KCl-LiCl salt to dissolve the fuel. This process results in a spent salt containing alkali, alkaline earth, lanthanide halides and small quantities of actinide halides, where the primary halide is chloride with a minor iodide fraction. Pacific Northwest National Laboratory (PNNL) is concurrently investigating two candidate waste forms for the Echem spent-salt: high-halide minerals (i.e., sodalite and cancrinite) and tellurite (TeO2)-based glasses. Both of these candidates showed promise in fiscal year (FY) 2009 and FY2010 with a simplified nonradioactive simulant of the Echem waste. Further testing was performed on these waste forms in FY2011 and FY2012 to assess the possibility of their use in a sustainable fuel cycle. This report summarizes the combined results from FY2011 and FY2012 efforts.

  14. HAW project. Demonstrative disposal of high-level radioactive wastes in the Asse salt mine

    International Nuclear Information System (INIS)

    Rothfuchs, T.; Duijves, K.; Stippler, R.

    1988-01-01

    Since 1968 the GSF has been carrying out research and development programs for the final disposal of high-level radioactive waste (HAW) in salt formations. The heat producing waste has been simulated so far by means of electrical heaters and also cobalt-60-sources. In order to improve the final concept for HAW disposal in salt formations the complete technical system of an underground repository is to be tested in an one-to-one scale test facility. To satisfy the test objectives thirty high radioactive canisters containing the radionuclides Cs-137 and Sr-90 will be emplaced in six boreholes located in two test galleries at the 800 m-level in the Asse salt mine. The duration of testing will be approximately five years. For the handling of the radioactive canisters and their emplacement into the boreholes a system consisting of transportation casks, transportation vehicle, disposal machine, and borehole slider will be developed and tested. The actual scientific investigation program is based on the estimation and observation of the interaction between the radioactive canisters and the rock salt. This program includes measurement of thermally and radiolytically induced water and gas release from the rock salt and the radiolytical decomposition of salt minerals. Also the thermally induced stress and deformation fields in the surrounding rock mass will be investigated carefully. The project is funded by the BMFT and the CEC and carrier out in close co-operation with the Netherlands Energy Research Foundation (ECN)

  15. The HAW project. Demonstrative disposal of high-level radioactive wastes in the Asse salt mine

    International Nuclear Information System (INIS)

    Rothfuchs, T.; Duijves, K.

    1988-04-01

    Since 1968 the GSF has been carrying out research and development programs for the final disposal of high-level radioactive waste (HAW) in salt formations. The heat producing waste has been simulated so far by means of electrical heaters and also cobalt-60-sources. In order to improve the final concept for HAW disposal in salt formations the complete technical system of an underground repository is to be tested in a one-to-one scale test facility. To satisfy the test objectives thirty high radioactive canisters containing the radionuclides Cs-137 and Sr-90 will be emplaced in six boreholes located in two test galleries at the 800 m-level in the Asse salt mine. The duration of testing will be approximately five years. For the handling of the radioactive canisters and their emplacement into the boreholes a system consisting of transportation casks, transportation vehicle, disposal machine, and borehole slider will be developed and tested. The actual scientific investigation program is based on the estimation and observation of the interaction between the radioactive canisters and the rock salt. This program includes measurement of thermally and radiolytically induced water and gas release from the rock salt and the radiolytical decomposition of salt minerals. Also the thermally induced stress and deformation fields in the surrounding rock mass will be investigated carefully. (orig./HP)

  16. Definition of the waste package environment for a repository located in salt

    International Nuclear Information System (INIS)

    Clark, D.E.; Bradley, D.J.

    1983-01-01

    The expected environmental conditions for emplaced waste packages in a salt repository are simulated in the materials testing program to evaluate performance. Synthetic brines, based on the analyses of actual brines (both intrusion and inclusion), are used for corrosion and leach testing. Elevated temperatures (to 150 0 C) and radiation fields of up to 10 3 rad/h are employed as conservative conditions to bracket expected performance and provide data for worst case scenarios. Obtaining a precise definition of the waste package environment in a salt repository and its change with time is closely tied to detailed site characterization of the candidate salt repository horizon. It is expected that field testing can augment some of the materials testing currently under way and can provide increased confidence in the predicted site-specific near-field conditions. 17 references, 5 figures, 1 table

  17. Possible salt mine sites for radioactive waste disposal in the northeastern states

    Energy Technology Data Exchange (ETDEWEB)

    Landes, K.K.

    1972-06-30

    The motivation for this investigation is the necessity for finding the safest possible repository for solid atomic plant wastes. It is believed that rooms mined in thick beds of salt would afford the best sanctuary. This is due especially to the impermeability of massive rock salt. This rock has enough plasticity so that it tends to give rather than fracture when disturbed by movements of the earth's crust. In addition, due to water conditions at the time of deposition, the rocks most commonly associated with salt (anhydrite and shale) are likewise relatively impervious. A number of areas have been selected for detailed discussion because of the excellence of the geological and environmental factors. The optimum requirements for a viable waste disposal prospect are described in detail and nine prospects are considered further.

  18. Possible salt mine sites for radioactive waste disposal in the northeastern states

    International Nuclear Information System (INIS)

    Landes, K.K.

    1972-01-01

    The motivation for this investigation is the necessity for finding the safest possible repository for solid atomic plant wastes. It is believed that rooms mined in thick beds of salt would afford the best sanctuary. This is due especially to the impermeability of massive rock salt. This rock has enough plasticity so that it tends to give rather than fracture when disturbed by movements of the earth's crust. In addition, due to water conditions at the time of deposition, the rocks most commonly associated with salt (anhydrite and shale) are likewise relatively impervious. A number of areas have been selected for detailed discussion because of the excellence of the geological and environmental factors. The optimum requirements for a viable waste disposal prospect are described in detail and nine prospects are considered further

  19. Computer simulation of an internally pressurized radioactive waste disposal room in a bedded salt formation

    International Nuclear Information System (INIS)

    Brown, W.T.; Weatherby, J.R.

    1991-01-01

    The Waste Isolation Pilot Plant (WIPP) near Carlsbad, New Mexico was created by the U.S. Department of Energy as an underground research and development facility to demonstrate the safe storage of transuranic waste generated from defense activities. This facility consists of storage rooms mined from a bedded salt formation at a depth of about 650 meters. Each room will accommodate about 6800 55-gallon drums filled with waste. After waste containers are emplaced, the storage rooms are to be backfilled with mined salt or other backfill materials. As time passes, reconsolidation of this backfill will reduce the hydraulic conductivity of the room. However, gases produced by decomposition and corrosion of waste and waste containers may cause a slow build-up of pressure which can retard consolidation of the waste and backfilled salt. The authors have developed a finite-element model of an idealized disposal room which is assumed to be perfectly sealed. The assumption that no gas escapes from the disposal room is a highly idealized and extreme condition which does not account for leakage paths, such as interbeds, that exist in the surrounding salt formation. This model has been used in a parametric study to determine how reconsolidation is influenced by various assumed gas generation rates and total amounts of gas generated. Results show that reductions in the gas generation, relative to the baseline case, can increase the degree of consolidation and reduce the peak gas pressure in disposal rooms. Even higher degrees of reconsolidation can be achieved by reducing both amounts and rates of gas generation. 8 refs., 4 figs., 1 tab

  20. Engineering study of the potential uses of salts from selective crystallization of Hanford tank wastes

    International Nuclear Information System (INIS)

    Hendrickson, D.W.

    1996-01-01

    The Clean Salt Process (CSP) is the fractional crystallization of nitrate salts from tank waste stored on the Hanford Site. This study reviews disposition options for a CSP product made from Hanford Site tank waste. These options range from public release to onsite low-level waste disposal to no action. Process, production, safety, environment, cost, schedule, and the amount of CSP material which may be used are factors considered in each option. The preferred alternative is offsite release of clean salt. Savings all be generated by excluding the material from low-level waste stabilization. Income would be received from sales of salt products. Savings and income from this alternative amount to $1,027 million, excluding the cost of CSP operations. Unless public sale of CSP products is approved, the material should be calcined. The carbonate form of the CSP could then be used as ballast in tank closure and stabilization efforts. Not including the cost of CSP operations, savings of $632 million would be realized. These savings would result from excluding the material from low-level waste stabilization and reducing purchases of chemicals for caustic recycle and stabilization and closure. Dose considerations for either alternative are favorable. No other cost-effective alternatives that were considered had the capacity to handle significant quantities of the CSP products. If CSP occurs, full-scale tank-waste stabilization could be done without building additional treatment facilities after Phase 1 (DOE 1996). Savings in capital and operating cost from this reduction in waste stabilization would be in addition to the other gains described

  1. Nuclear waste repository simulation experiments. Asse salt mine: Annual report 1984

    International Nuclear Information System (INIS)

    Rothfuchs, T.; Feddersen, H.K.; Schwarzianeck, P.; Staupendahl, G.; Coyle, A.J.; Kalia, H.; Eckert, J.

    1985-01-01

    This is the Second Annual Report (1984) which describes experiments simulating a nuclear waste repository at the 800 meter-level of the Asse Salt Mine in the Federal Republic of Germany. The report describes the Asse Salt Mine, the test equipment, and the pretest properties of the salt in the mine and in the vicinity of the test area. Also included are test data for the first sixteen months of operation on the following: brine migration rates, thermal mechanical behavior of the salt (including room closure, stress readings and thermal profiles) and borehole gas pressures. In addition to field data laboratory analyses of results are also included in this report. The duration of the experiment will be two years, ending in December 1985. (orig.)

  2. Nuclear waste repository simulation experiments, Asse salt mine, Federal Republic of Germany. Annual report 1984

    International Nuclear Information System (INIS)

    Rothfuchs, T.; Feddersen, H.K.; Schwarzianeck, P.; Staupendahl, G.; Coyle, A.J.; Eckert, J.; Kalia, H.

    1986-07-01

    This is the second joint annual report (1984) on experiments simulating a nuclear waste repository at the 800-m (2624-ft) level of the Asse salt mine in the Federal Republic of Germany. This report describes the Asse salt mine, the test equipment, and the pretest properties of the salt in the mine and in the vicinity of the test area. Also included are test data for the first 19 months of operation on the following: brine migration rates, thermal mechanical behavior of the salt (including room closure, stress reading, and thermal profiles), and borehole gas pressures. In addition to field data, laboratory analyses of results are included in this report. The duration of the experiment will be 2 years, ending in December 1985

  3. Radioactive waste isolation in salt: geochemistry of brine in rock salt in temperature gradients and gamma-radiation fields - a selective annotated bibliography

    International Nuclear Information System (INIS)

    Hull, A.B.; Williams, L.B.

    1985-07-01

    Evaluation of the extensive research concerning brine geochemistry and transport is critically important to successful exploitation of a salt formation for isolating high-level radioactive waste. This annotated bibliography has been compiled from documents considered to provide classic background material on the interactions between brine and rock salt, as well as the most important results from more recent research. Each summary elucidates the information or data most pertinent to situations encountered in siting, constructing, and operating a mined repository in salt for high-level radioactive waste. The research topics covered include the basic geology, depositional environment, mineralogy, and structure of evaporite and domal salts, as well as fluid inclusions, brine chemistry, thermal and gamma-radiation effects, radionuclide migration, and thermodynamic properties of salts and brines. 4 figs., 6 tabs

  4. Immobilization of LiCl-Li 2 O pyroprocessing salt wastes in chlorosodalite using glass-bonded hydrothermal and salt-occlusion methods

    Energy Technology Data Exchange (ETDEWEB)

    Riley, Brian J.; Peterson, Jacob A.; Kroll, Jared O.; Frank, Steven M.

    2018-04-01

    In this study, salt occlusion and hydrothermal processes were used to make chlorosodalite through reaction with a high-LiCl salt simulating a waste stream following pyrochemical treatment of oxide-based used nuclear fuel. Some products were reacted with glass binders to increase chlorosodalite yield through alkali ion exchange and aide in densification. Hydrothermal processes included reaction of the salt simulant in an acid digestion vessel with either zeolite 4A or sodium aluminate and colloidal silica. Chlorosodalite yields in the crystalline products were nearly complete in the glass-bonded materials at values of 100 mass% for the salt-occlusion method, up to 99.0 mass% for the hydrothermal synthesis with zeolite 4A, and up to 96 mass% for the hydrothermal synthesis with sodium aluminate and colloidal silica. These results show promise for using chemically stable chlorosodalite to immobilize oxide reduction salt wastes.

  5. Immobilization of LiCl-Li2O pyroprocessing salt wastes in chlorosodalite using glass-bonded hydrothermal and salt-occlusion methods

    Science.gov (United States)

    Riley, Brian J.; Peterson, Jacob A.; Kroll, Jared O.; Frank, Steven M.

    2018-04-01

    In this study, hydrothermal and salt-occlusion processes were used to make chlorosodalite through reactions with a high-LiCl salt simulating a waste stream generated from pyrochemical treatment of oxide-based used nuclear fuel. Some products were reacted with glass binders to increase chlorosodalite yield through alkali ion exchange and to aid in densification. Hydrothermal processes included reaction of the salt simulant in an autoclave with either zeolite 4A or sodium aluminate and colloidal silica. Chlorosodalite yields in the crystalline products were nearly complete in the glass-bonded materials at values of 100 mass% for the salt-occlusion method, up to 99.0 mass% for the hydrothermal synthesis with zeolite 4A, and up to 96 mass% for the hydrothermal synthesis with sodium aluminate and colloidal silica. These results show promise for using chemically stable chlorosodalite to immobilize oxide reduction salt wastes.

  6. Radioactive waste isolation in salt: geochemistry of brine in rock salt in temperature gradients and gamma-radiation fields - a selective annotated bibliography

    Energy Technology Data Exchange (ETDEWEB)

    Hull, A.B.; Williams, L.B.

    1985-07-01

    Evaluation of the extensive research concerning brine geochemistry and transport is critically important to successful exploitation of a salt formation for isolating high-level radioactive waste. This annotated bibliography has been compiled from documents considered to provide classic background material on the interactions between brine and rock salt, as well as the most important results from more recent research. Each summary elucidates the information or data most pertinent to situations encountered in siting, constructing, and operating a mined repository in salt for high-level radioactive waste. The research topics covered include the basic geology, depositional environment, mineralogy, and structure of evaporite and domal salts, as well as fluid inclusions, brine chemistry, thermal and gamma-radiation effects, radionuclide migration, and thermodynamic properties of salts and brines. 4 figs., 6 tabs.

  7. Chemical modeling of nuclear waste repositories in the salt repository project

    International Nuclear Information System (INIS)

    Jansen, G.; Raines, G.E.; Kircher, J.F.; Hubbard, N.

    1985-01-01

    Salt deposits contain small amounts of water as brine in fluid inclusions in halite and in hydrous minerals, e.g., clays, kieserite (MgSO 4 . H 2 O) and carnallite (KMgCl 3 . 6H 2 O). For the candidate salt deposits, the total amounts of water as volume % brine are: Palo Duro Basin, Texas, approximately 1.8; Paradox Basin, Utah, approximately 5.0 for the carnallite-marker zone, and less than approximately 0.5 below this zone; Gulf Coast salt domes, less than 0.15. For the Palo Duro and Paradox salt, the brines are Mg-rich (approximately 20,000 mg/L to approximately 100,000 mg/L) and sometimes Ca-rich (up to about 20,000 mg/L) NaCl brines. Brine migration calculations have been made using calculations of the time-variant thermal gradient around the waste packages and conservatively high brine volumes in the salt (5.0 volume % for the Texas and Utah sites and 0.5 volume % for the Gulf Coast) as input data. The maximum amounts of brine that eventually migrate to each waste package are about 1.0m 3 (for 5.0 volume % brine) and 0.2m 3 (for 0.5 volume % brine). With current conceptual designs for waste package overpacks (10 to 15 cm thick low-carbon steel), the waste package is not breached by uniform corrosion within 10,000 years. In brines this material thus far shows only uniform corrosion. For the expected conditions, where the brine is provided solely by brine migration, the brine is consumed by reaction with the iron of the overpack nearly as fast as it migrates to the waste package. Therefore, for the expected conditions, data about corrosion rates, radiolysis, etc., are not important. However, it is essential that accurate volumes of in-migrating brine can be calculated

  8. The advantages of a salt/bentonite backfill for Waste Isolation Pilot Plant disposal rooms

    International Nuclear Information System (INIS)

    Butcher, B.M.; Novak, C.F.; Jercinovic, M.

    1991-04-01

    A 70/30 wt% salt/bentonite mixture is shown to be preferable to pure crushed salt as backfill for disposal rooms in the Waste Isolation Pilot Plant (WIPP). This report discusses several selection criteria used to arrive at this conclusion: the need for low permeability and porosity after closure, chemical stability with the surroundings, adequate strength to avoid shear erosion from human intrusion, ease of emplacement, and sorption potential for brine and radionuclides. Both salt and salt/bentonite are expected to consolidate to a final state of impermeability (i.e., ≤ 10 -18 m 2 ) adequate for satisfying federal nuclear regulations. Any advantage of the salt/bentonite mixture is dependent upon bentonite's potential for sorbing brine and radionuclides. Estimates suggest that bentonite's sorption potential for water in brine is much less than for pure water. While no credit is presently taken for brine sorption in salt/bentonite backfill, the possibility that some amount of inflowing brine would be chemically bound is considered likely. Bentonite may also sorb much of the plutonium, americium, and neptunium within the disposal room inventory. Sorption would be effective only if a major portion of the backfill is in contact with radioactive brine. Brine flow from the waste out through highly localized channels in the backfill would negate sorption effectiveness. Although the sorption potentials of bentonite for both brine and radionuclides are not ideal, they are distinctly beneficial. Furthermore, no detrimental aspects of adding bentonite to the salt as a backfill have been identified. These two observations are the major reasons for selecting salt/bentonite as a backfill within the WIPP. 39 refs., 16 figs., 6 tabs

  9. Options Assessment Report: Treatment of Nitrate Salt Waste at Los Alamos National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Robinson, Bruce Alan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stevens, Patrice Ann [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-12-17

    This report documents the methodology used to select a method of treatment for the remediated nitrate salt (RNS) and unremediated nitrate salt (UNS) waste containers at Los Alamos National Laboratory (LANL). The method selected should treat the containerized waste in a manner that renders the waste safe and suitable for transport and final disposal in the Waste Isolation Pilot Plant (WIPP) repository, under specifications listed in the WIPP Waste Acceptance Criteria (DOE/CBFO, 2013). LANL recognizes that the results must be thoroughly vetted with the New Mexico Environment Department (NMED) and that a modification to the LANL Hazardous Waste Facility Permit is a necessary step before implementation of this or any treatment option. Likewise, facility readiness and safety basis approvals must be received from the Department of Energy (DOE). This report presents LANL’s preferred option, and the documentation of the process for reaching the recommended treatment option for RNS and UNS waste, and is presented for consideration by NMED and DOE.

  10. Options assessment report: Treatment of nitrate salt waste at Los Alamos National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Robinson, Bruce Alan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stevens, Patrice Ann [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-09-16

    This report documents the methodology used to select a method of treatment for the remediated nitrate salt (RNS) and unremediated nitrate salt (UNS) waste containers at Los Alamos National Laboratory (LANL). The method selected should treat the containerized waste in a manner that renders the waste safe and suitable for transport and final disposal in the Waste Isolation Pilot Plant (WIPP) repository, under specifications listed in the WIPP Waste Acceptance Criteria (DOE/CBFO, 2013). LANL recognized that the results must be thoroughly vetted with the New Mexico Environment Department (NMED) and the a modification to the LANL Hazardous Waste Facility Permit is a necessary step before implementation of this or any treatment option. Likewise, facility readiness and safety basis approvals must be received from the Department of Energy (DOE). This report presents LANL's preferred option, and the documentation of the process for reaching the recommended treatment option for RNS and UNS waste, and is presented for consideration by NMED and DOE.

  11. Waste package materials testing for a salt repository: 1983 status summary report

    International Nuclear Information System (INIS)

    Moak, D.P.

    1986-09-01

    The United States plans to safely dispose of nuclear waste in deep, stable geologic formations. As part of these plans, the US Department of Energy is sponsoring research on the designing and testing of waste packages and waste package materials. This fiscal year 1983 status report summarizes recent results of waste package materials testing in a salt environment. The results from these tests will be used by waste package designers and performance assessment experts. Release characteristics data are available on two waste forms (spent fuel and waste-containing glass) that were exposed to leaching tests at various radiation levels, temperatures, pH, glass surface area to solution volume ratios, and brine solutions simulating expected salt repository conditions. Candidate materials tested for corrosion resistance and other properties include iron alloys; TI-CODE 12, the most promising titanium alloy for containment; and nickel alloys. In component interaction testing, synergistic effects have not ruled out any candidate material. 21 refs., 37 figs., 15 tabs

  12. X-ray diffraction of slag-based sodium salt waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Langton, C. A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Missimer, D. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2014-09-30

    The attached report documents sample preparation and x-ray diffraction results for a series of cement and blended cement matrices prepared with either water or a 4.4 M Na salt solution. The objective of the study was to provide initial phase characterization for the Cementitious Barriers Partnership reference case cementitious salt waste form. This information can be used to: 1) generate a base line for the evolution of the waste form as a function of time and conditions, 2) potentially to design new binders based on mineralogy of the binder, 3) understand and predict anion and cation leaching behavior of contaminants of concern, and 4) predict performance of the waste forms for which phase solubility and thermodynamic data are available.

  13. Recovery of plutonium and americium from chloride salt wastes by solvent extraction

    International Nuclear Information System (INIS)

    Reichley-Yinger, L.; Vandegrift, G.F.

    1987-01-01

    Plutonium and americium can be recovered from aqueous waste solutions containing a mixture of HCl and chloride salt wastes by the coupling of two solvent extraction systems: tributyl phosphate (TBP) in tetrachloroethylene (TCE) and octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO) in TCE. In the flowsheet developed, the salt wastes are dissolved in HCl, the Pu(III) is oxidized to the IV state with NaClO 2 and recovered in the TBP-TCE cycle, and the Am is then removed from the resultant raffinate by the CMPO-TCE cycle. The consequences of the feed solution composition and extraction behavior of these species on the process flowsheet design, the Pu-product purity, and the decontamination of the aqueous raffinate from transuranic elements are discussed. 16 refs., 6 figs

  14. FGF-23 in fibrous dysplasia of bone and its relationship to renal phosphate wasting

    Science.gov (United States)

    Riminucci, Mara; Collins, Michael T.; Fedarko, Neal S.; Cherman, Natasha; Corsi, Alessandro; White, Kenneth E.; Waguespack, Steven; Gupta, Anurag; Hannon, Tamara; Econs, Michael J.; Bianco, Paolo; Gehron Robey, Pamela

    2003-01-01

    FGF-23, a novel member of the FGF family, is the product of the gene mutated in autosomal dominant hypophosphatemic rickets (ADHR). FGF-23 has been proposed as a circulating factor causing renal phosphate wasting not only in ADHR (as a result of inadequate degradation), but also in tumor-induced osteomalacia (as a result of excess synthesis by tumor cells). Renal phosphate wasting occurs in approximately 50% of patients with McCune-Albright syndrome (MAS) and fibrous dysplasia of bone (FD), which result from postzygotic mutations of the GNAS1 gene. We found that FGF-23 is produced by normal and FD osteoprogenitors and bone-forming cells in vivo and in vitro. In situ hybridization analysis of FGF-23 mRNA expression identified “fibrous” cells, osteogenic cells, and cells associated with microvascular walls as specific cellular sources of FGF-23 in FD. Serum levels of FGF-23 were increased in FD/MAS patients compared with normal age-matched controls and significantly higher in FD/MAS patients with renal phosphate wasting compared with those without, and correlated with disease burden bone turnover markers commonly used to assess disease activity. Production of FGF-23 by FD tissue may play an important role in the renal phosphate–wasting syndrome associated with FD/MAS. PMID:12952917

  15. Subsurface geology of a potential waste emplacement site, Salt Valley Anticline, Grand County, Utah

    Science.gov (United States)

    Hite, R.J.

    1977-01-01

    The Salt Valley anticline, which is located about 32 km northeast of Moab, Utah, is perhaps one of the most favorable waste emplacement sites in the Paradox basin. The site, which includes about 7.8 km 2, is highly accessible and is adjacent to a railroad. The anticline is one of a series of northwest-trending salt anticlines lying along the northeast edge of the Paradox basin. These anticlines are cored by evaporites of the Paradox Member of the Hermosa Formation of Middle Pennsylvanian age. The central core of the Salt Valley anticline forms a ridgelike mass of evaporites that has an estimated amplitude of 3,600 m. The evaporite core consists of about 87 percent halite rock, which includes some potash deposits; the remainder is black shale, silty dolomite, and anhydrite. The latter three lithologies are referred to as 'marker beds.' Using geophysical logs from drill holes on the anticline, it is possible to demonstrate that the marker beds are complexly folded and faulted. Available data concerning the geothermal gradient and heatflow at the site indicate that heat from emplaced wastes should be rapidly dissipated. Potentially exploitable resources of potash and petroleum are present at Salt Valley. Development of these resources may conflict with use of the site for waste emplacement.

  16. Subsurface geology of a potential waste emplacement site, Salt Valley Anticline, Grand County, Utah

    International Nuclear Information System (INIS)

    Hite, R.J.

    1977-01-01

    The Salt Valley anticline, which is located about 32 km northeast of Moab, Utah, is perhaps one of the most favorable waste emplacement sites in the Paradox basin. The site, which includes about 7.8 km 2 , is highly accessible and is adjacent to a railroad. The anticline is one of a series of northwest-trending salt antilcines lying along the northeast edge of the Paradox basin. These anticlines are cored by evaporites of the Paradox Member of the Hermosa Formation of Middle Pennsylvanian age. The central core of the Salt Valley anticline forms a ridgelike mass of evaporites that has an estimated amplitude of 3,600 m. The evaporite core consists of about 87 percent halite rock, which includes some potash deposits; the remainder is black shale, silty dolomite, and anhydrite. The latter three lithologies are referred to as ''marker beds.'' Using geophysical logs from drill holes on the anticline, it is possible to demonstrate that the marker beds are complexly folded and faulted. Available data concerning the geothermal gradient and heatflow at the site indicate that heat from emplaced wastes should be rapidly dissipated. Potentially exploitable resources of potash and petroleum are present at Salt Valley. Development of these resources may conflict with use of the site for waste emplacement

  17. Temperature distributions in a salt formation used for the ultimate disposal of high-level radioactive waste

    International Nuclear Information System (INIS)

    Ploumen, P.

    1980-01-01

    In the Federal Republic of Germany the works on waste disposal is focussed on the utilization of a salt formation for ultimate disposal of radioactive wastes. Heat released from the high-level waste will be dissipated in the salt and the surrounding geologic formations. The occuring temperature distributions will be calculated with computer codes. A survey of the developed computer codes will be shown; the results for a selected example, taking into account the loading sequence of the waste, the mine ventilation as well as an air gap between the waste and the salt, will be discussed. Furthermore it will be shown that by varying the disposal parameters, the maximum salt temperature can be below any described value. (Auth.)

  18. Asse salt mine nuclear waste repository simulation experiments

    International Nuclear Information System (INIS)

    Coyle, A.J.

    1983-01-01

    The field tests underway in Asse, Federal Republic of Germany are dicected toward the development of test plans, techniques and equipment to be used in Exploratory Shafts or At Depth Test Facilities confirmation tests. These simulated repository tests will also provide information which address the following issues: brine migration (liquid and vapor); radiation effects of gamma rays; gas generation caused by radiation and corrosion; accelerated corrosion and leaching; altered properties of salt (the effects of heat, radiation and brine); and the effects of heat and radiation on test assemblies, instruments, and various materials exposed to repository conditions. This paper is a status of the first 82 days of operation of the Asse Brine Migration Tests, which were initiated on May 25, 1983. 6 references

  19. Waste segregation analysis for salt well pumping in the 200 W Area -- Task 3.4

    International Nuclear Information System (INIS)

    Reynolds, D.A.

    1995-01-01

    There is an estimated 7 million liters (1.9 million gallons) of potentially complexed waste that need to be pumped from single-shell tanks (SST) in the 200 West Area. This represents up to 40% of the salt well liquor that needs to be pumped in the 200 West Area. There are three double-shell (DST) tanks in the 241-SY tank farm in the 200 West Area. Tank 241-SY-101 is full and not usable. Tank 241-SY-102 has a transuranic (TRU) sludge in the bottom. Current rules prohibit mixing complexed waste with TRU waste. Tank 241-SY-103 has three major problems. First, 241-SY-103 is on the Flammable Watch list. Second, adding waste to tank 241-SY-103 has the potential for an episodic release of hydrogen gas. Third, 241-SY-103 will not hold all of the potentially complexed waste from the SSTs. This document looks at more details regarding the salt well pumping of the 200 West Area tank farm. Some options are considered but it is beyond the scope of this document to provide an in-depth study necessary to provide a defensible solution to the complexed waste problem

  20. Corrosion of candidate iron-base waste package structural barrier materials in moist salt environments

    International Nuclear Information System (INIS)

    Westerman, R.E.; Pitman, S.G.

    1984-11-01

    Mild steels are considered to be strong candidates for waste package structural barrier (e.g., overpack) applications in salt repositories. Corrosion rates of these materials determined in autoclave tests utilizing a simulated intrusion brine based on Permian Basin core samples are low, generally <25 μm (1 mil) per year. When the steels are exposed to moist salts containing simulated inclusion brines, the corrosion rates are found to increase significantly. The magnesium in the inclusion brine component of the environment is believed to be responsible for the increased corrosion rates. 1 reference, 4 figures, 2 tables

  1. Evaluation of iron-base materials for waste package containers in a salt repository

    International Nuclear Information System (INIS)

    Westerman, R.E.; Nelson, J.L.; Kuhn, W.L.; Basham, S.G.; Moak, D.A.; Pitman, S.G.

    1983-11-01

    Design studies for high-level nuclear waste packages for salt repositories have identified low-carbon steel as a candidate material for containers. Among the requirements are strength, corrosion resistance, and fabricability. The studies of the corrosion resistance and structural stability of iron-base materials (particularly low-carbon steel) are treated in this paper. The materials have been exposed in brines that are characteristic of the potential sites for salt repositories. The effects of temperature, radiation level, oxygen level and other parameters are under investigation. The initial development of corrosion models for these environments is presented with discussion of the key mechanisms under consideration. 6 references, 5 figures

  2. Hydrous mineral dehydration around heat-generating nuclear waste in bedded salt formations.

    Science.gov (United States)

    Jordan, Amy B; Boukhalfa, Hakim; Caporuscio, Florie A; Robinson, Bruce A; Stauffer, Philip H

    2015-06-02

    Heat-generating nuclear waste disposal in bedded salt during the first two years after waste emplacement is explored using numerical simulations tied to experiments of hydrous mineral dehydration. Heating impure salt samples to temperatures of 265 °C can release over 20% by mass of hydrous minerals as water. Three steps in a series of dehydration reactions are measured (65, 110, and 265 °C), and water loss associated with each step is averaged from experimental data into a water source model. Simulations using this dehydration model are used to predict temperature, moisture, and porosity after heating by 750-W waste canisters, assuming hydrous mineral mass fractions from 0 to 10%. The formation of a three-phase heat pipe (with counter-circulation of vapor and brine) occurs as water vapor is driven away from the heat source, condenses, and flows back toward the heat source, leading to changes in porosity, permeability, temperature, saturation, and thermal conductivity of the backfill salt surrounding the waste canisters. Heat pipe formation depends on temperature, moisture availability, and mobility. In certain cases, dehydration of hydrous minerals provides sufficient extra moisture to push the system into a sustained heat pipe, where simulations neglecting this process do not.

  3. Molten salt oxidation of mixed wastes: Separation of radioactive materials and Resource Conservation and Recovery Act (RCRA) materials

    International Nuclear Information System (INIS)

    Bell, J.T.; Haas, P.A.; Rudolph, J.C.

    1995-01-01

    The Oak Ridge National Laboratory (ORNL) is participating in a program to apply a molten salt oxidation (MSO) process to treatment of mixed (radioactive and RCRA) wastes. The salt residues from the MSO treatment will require further separations or other processing to prepare them for final disposal. A bench-scale MSO apparatus is being installed at ORNL and will be operated on real Oak Ridge wastes. The treatment concepts to be tested and demonstrated on the salt residues from real wastes are described

  4. Equipment evaluation for low density polyethylene encapsulated nitrate salt waste at the Rocky Flats Plant

    International Nuclear Information System (INIS)

    Yamada, W.I.; Faucette, A.M.; Jantzen, R.C.; Logsdon, B.W.; Oldham, J.H.; Saiki, D.M.; Yudnich, R.J.

    1993-01-01

    Mixed wastes at the Rocky Flats Plant (RFP) are subject to regulation by the Resource Conservation and Recovery Act (RCRA). Polymer solidification is being developed as a final treatment technology for several of these mixed wastes, including nitrate salts. Encapsulation nitrate salts with low density polyethylene (LDPE) has been the preliminary focus of the RFP polymer solidification effort. Literature reviews, industry surveys, and lab-scale and pilot-scale tests have been conducted to evaluate several options for encapsulating nitrate salts with LDPE. Most of the effort has focused on identifying compatible drying and extrusion technologies. Other processing options, specifically meltration and non-heated compounding machines, were also investigated. The best approach appears to be pretreatment of the nitrate salt waste brine in either a vertical or horizontal thin film evaporator followed by compounding of the dried waste with LDPE in an intermeshing, co-rotating, twin-screw extruder. Additional pilot-scale tests planned for the fall of 1993 should further support this recommendation. Preliminary evaluation work indicates that meltration is not possible at atmospheric pressure with the LDPE (Chevron PE-1409) provided by RFP. However, meltration should be possible at atmospheric pressure using another LDPE formulation with altered physical and rheological properties: Lower molecular weight and lower viscosity (Epoline C-15). Contract modifications are now in process to allow a follow-on pilot scale demonstration. Questions regarding changed safety and physical properties of the resultant LDPE waste form due to use of the Epoline C-15 will be addressed. No additional work with non-heated mixer compounder machines is planned at this time

  5. Test Results and Comparison of Triaxial Strength Testing of Waste Isolation Pilot Plant Clean Salt

    Energy Technology Data Exchange (ETDEWEB)

    Buchholz, Stuart A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-12-01

    This memorandum documents laboratory thermomechanical triaxial strength testing of Waste Isolation Pilot Plant (WIPP) clean salt. The limited study completed independent, adjunct laboratory tests in the United States to assist in validating similar testing results being provided by the German facilities. The testing protocol consisted of completing confined triaxial, constant strain rate strength tests of intact WIPP clean salt at temperatures of 25°C and 100°C and at multiple confining pressures. The stratigraphy at WIPP also includes salt that has been labeled “argillaceous.” The much larger test matrix conducted in Germany included both the so-called clean and argillaceous salts. When combined, the total database of laboratory results will be used to develop input parameters for models, assess adequacy of existing models, and predict material behavior. These laboratory studies are also consistent with the goals of the international salt repository research program. The goal of this study was to complete a subset of a test matrix on clean salt from the WIPP undertaken by German research groups. The work was performed at RESPEC in Rapid City, South Dakota. A rigorous Quality Assurance protocol was applied, such that corroboration provides the potential of qualifying all of the test data gathered by German research groups.

  6. Considerations of the Differences between Bedded and Domal Salt Pertaining to Disposal of Heat-Generating Nuclear Waste

    International Nuclear Information System (INIS)

    Hansen, Francis D.; Kuhlman, Kristopher L.; Sobolik, Steven R.

    2016-01-01

    Salt formations hold promise for eternal removal of nuclear waste from our biosphere. Germany and the United States have ample salt formations for this purpose, ranging from flat-bedded formations to geologically mature dome structures. As both nations revisit nuclear waste disposal options, the choice between bedded, domal, or intermediate pillow formations is once again a contemporary issue. For decades, favorable attributes of salt as a disposal medium have been extoled and evaluated, carefully and thoroughly. Yet, a sense of discovery continues as science and engineering interrogate naturally heterogeneous systems. Salt formations are impermeable to fluids. Excavation-induced fractures heal as seal systems are placed or natural closure progresses toward equilibrium. Engineering required for nuclear waste disposal gains from mining and storage industries, as humans have been mining salt for millennia. This great intellectual warehouse has been honed and distilled, but not perfected, for all nuances of nuclear waste disposal. Nonetheless, nations are able and have already produced suitable license applications for radioactive waste disposal in salt. A remaining conundrum is site location. Salt formations provide isolation, and geotechnical barriers reestablish impermeability after waste is placed in the geology. Between excavation and closure, physical, mechanical, thermal, chemical, and hydrological processes ensue. Positive attributes for isolation in salt have many commonalities independent of the geologic setting. In some cases, specific details of the environment will affect the disposal concept and thereby define interaction of features, events and processes, while simultaneously influencing scenario development. Here we identify and discuss high-level differences and similarities of bedded and domal salt formations. Positive geologic and engineering attributes for disposal purposes are more common among salt formations than are significant differences

  7. Considerations of the Differences between Bedded and Domal Salt Pertaining to Disposal of Heat-Generating Nuclear Waste

    Energy Technology Data Exchange (ETDEWEB)

    Hansen, Francis D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Kuhlman, Kristopher L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sobolik, Steven R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-07-07

    Salt formations hold promise for eternal removal of nuclear waste from our biosphere. Germany and the United States have ample salt formations for this purpose, ranging from flat-bedded formations to geologically mature dome structures. As both nations revisit nuclear waste disposal options, the choice between bedded, domal, or intermediate pillow formations is once again a contemporary issue. For decades, favorable attributes of salt as a disposal medium have been extoled and evaluated, carefully and thoroughly. Yet, a sense of discovery continues as science and engineering interrogate naturally heterogeneous systems. Salt formations are impermeable to fluids. Excavation-induced fractures heal as seal systems are placed or natural closure progresses toward equilibrium. Engineering required for nuclear waste disposal gains from mining and storage industries, as humans have been mining salt for millennia. This great intellectual warehouse has been honed and distilled, but not perfected, for all nuances of nuclear waste disposal. Nonetheless, nations are able and have already produced suitable license applications for radioactive waste disposal in salt. A remaining conundrum is site location. Salt formations provide isolation and geotechnical barriers reestablish impermeability after waste is placed in the geology. Between excavation and closure, physical, mechanical, thermal, chemical, and hydrological processes ensue. Positive attributes for isolation in salt have many commonalities independent of the geologic setting. In some cases, specific details of the environment will affect the disposal concept and thereby define interaction of features, events and processes, while simultaneously influencing scenario development. Here we identify and discuss high-level differences and similarities of bedded and domal salt formations. Positive geologic and engineering attributes for disposal purposes are more common among salt formations than are significant differences

  8. Considerations of the Differences between Bedded and Domal Salt Pertaining to Disposal of Heat-Generating Nuclear Waste

    Energy Technology Data Exchange (ETDEWEB)

    Hansen, Francis D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Kuhlman, Kristopher L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Sobolik, Steven R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-07-07

    Salt formations hold promise for eternal removal of nuclear waste from our biosphere. Germany and the United States have ample salt formations for this purpose, ranging from flat-bedded formations to geologically mature dome structures. As both nations revisit nuclear waste disposal options, the choice between bedded, domal, or intermediate pillow formations is once again a contemporary issue. For decades, favorable attributes of salt as a disposal medium have been extoled and evaluated, carefully and thoroughly. Yet, a sense of discovery continues as science and engineering interrogate naturally heterogeneous systems. Salt formations are impermeable to fluids. Excavation-induced fractures heal as seal systems are placed or natural closure progresses toward equilibrium. Engineering required for nuclear waste disposal gains from mining and storage industries, as humans have been mining salt for millennia. This great intellectual warehouse has been honed and distilled, but not perfected, for all nuances of nuclear waste disposal. Nonetheless, nations are able and have already produced suitable license applications for radioactive waste disposal in salt. A remaining conundrum is site location. Salt formations provide isolation, and geotechnical barriers reestablish impermeability after waste is placed in the geology. Between excavation and closure, physical, mechanical, thermal, chemical, and hydrological processes ensue. Positive attributes for isolation in salt have many commonalities independent of the geologic setting. In some cases, specific details of the environment will affect the disposal concept and thereby define interaction of features, events and processes, while simultaneously influencing scenario development. Here we identify and discuss high-level differences and similarities of bedded and domal salt formations. Positive geologic and engineering attributes for disposal purposes are more common among salt formations than are significant differences

  9. Modeling Solute Thermokinetics in LiCI-KCI Molten Salt for Nuclear Waste Separation

    Energy Technology Data Exchange (ETDEWEB)

    Morgan, Dane; Eapen, Jacob

    2013-10-01

    Recovery of actinides is an integral part of a closed nuclear fuel cycle. Pyrometallurgical nuclear fuel recycling processes have been developed in the past for recovering actinides from spent metallic and nitride fuels. The process is essentially to dissolve the spent fuel in a molten salt and then extract just the actinides for reuse in a reactor. Extraction is typically done through electrorefining, which involves electrochemical reduction of the dissolved actinides and plating onto a cathode. Knowledge of a number of basic thermokinetic properties of salts and salt-fuel mixtures is necessary for optimizing present and developing new approaches for pyrometallurgical waste processing. The properties of salt-fuel mixtures are presently being studied, but there are so many solutes and varying concentrations that direct experimental investigation is prohibitively time consuming and expensive (particularly for radioactive elements like Pu). Therefore, there is a need to reduce the number of required experiments through modeling of salt and salt-fuel mixture properties. This project will develop first-principles-based molecular modeling and simulation approaches to predict fundamental thermokinetic properties of dissolved actinides and fission products in molten salts. The focus of the proposed work is on property changes with higher concentrations (up to 5 mol%) of dissolved fuel components, where there is still very limited experimental data. The properties predicted with the modeling will be density, which is used to assess the amount of dissolved material in the salt; diffusion coefficients, which can control rates of material transport during separation; and solute activity, which determines total solubility and reduction potentials used during electrorefining. The work will focus on La, Sr, and U, which are chosen to include the important distinct categories of lanthanides, alkali earths, and actinides, respectively. Studies will be performed using LiCl-KCl salt

  10. Extraction, scrub, and strip test results for the solvent transfer to salt waste processing facility

    Energy Technology Data Exchange (ETDEWEB)

    Peters, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-09-07

    The Savannah River National Laboratory (SRNL) prepared approximately 240 gallons of Caustic-Side Solvent Extraction (CSSX) solvent for use at the Salt Waste Processing Facility (SWPF). An Extraction, Scrub, and Strip (ESS) test was performed on a sample of the prepared solvent using a salt solution prepared by Parsons to determine cesium distribution ratios (D(Cs)), and cesium concentration in the strip effluent (SE) and decontaminated salt solution (DSS) streams. This data will be used by Parsons to help qualify the solvent for use at the SWPF. The ESS test showed acceptable performance of the solvent for extraction, scrub, and strip operations. The extraction D(Cs) measured 15.5, exceeding the required value of 8. This value is consistent with results from previous ESS tests using similar solvent formulations. Similarly, scrub and strip cesium distribution ratios fell within acceptable ranges.

  11. The dispersal and impact of salt from surface storage piles the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    Reith, C.C.; Louderbough, E.T.

    1986-01-01

    A comprehensive program of ecological studies occurs at the Waste Isolation Pilot Plant (WIPP) in an effort to detect and quantify impacts of excavated salt which is stored on the surface in two piles: one having originated in 1980, the other in 1984. Both piles are surrounded by berms which channel runoff to holding ponds, so nearly all dispersal is due to the resuspension, transport, and deposition of salt particles by wind. Ecological parameters which have been monitored since 1984 include: visual evidence (via photography), soil properties, microbial activity, leaf-litter decomposition, seedling emergence, plant foliar cover, and plant species diversity. These are periodically assessed at experimental plots near the salt piles, and at control plots several kilometers away

  12. Radiant energy dissipation during final storage of high-level radioactive waste in rock salt

    International Nuclear Information System (INIS)

    Ramthun, H.

    1981-08-01

    A final disposal concept is assumed where the high-active waste from 1400 t of uranium, remaining after conditioning, is solidified in borosilicate glass and distributed in 1.760 waste casks. These containers 1.2 m in height and 0.3 m in diameter are to be buried 10 years after the fuel is removed from the reactor in the 300 m deep boreholes of a salt dome. For this design the mean absorbed dose rates are calculated in the glass die (3.9 Gy/s), the steel mantle (0.26 Gy/s) and in the salt rock (0.12 Gy/s at a distance of 1 cm and 0.034 Gy/s at a distance of 9 cm from the container surface) valid at the beginning of disposal. The risk involved with these amounts of stored lattice energy is shortly discussed. (orig.) [de

  13. From cerebral salt wasting to diabetes insipidus with adipsia: case report of a child with craniopharyngioma.

    Science.gov (United States)

    Raghunathan, Veena; Dhaliwal, Maninder Singh; Gupta, Aditya; Jevalikar, Ganesh

    2015-03-01

    Craniopharyngioma is associated with a wide and interesting variety of sodium states both by itself and following surgical resection. These are often challenging to diagnose, especially given their dynamic nature during the perioperative course. We present the case of a boy with craniopharyngioma who had hyponatremia due to cerebral salt wasting preoperatively, developed diabetes insipidus (DI) intraoperatively and proceeded to develop hypernatremia with adipsic DI. Cerebral salt wasting is a rare presenting feature of craniopharyngioma. Postoperative DI can be associated with thirst abnormalities including adipsia due to hypothalamic damage; careful monitoring and a high index of suspicion are required for its detection. Adipsic DI is a difficult condition to manage; hence a conservative surgical approach is suggested.

  14. Phase Equilibrium Studies of Savannah River Tanks and Feed Streams for the Salt Waste Processing Facility

    Energy Technology Data Exchange (ETDEWEB)

    Weber, C.F.

    2001-06-19

    A chemical equilibrium model is developed and used to evaluate supersaturation of tanks and proposed feed streams to the Salt Waste Processing Facility. The model uses Pitzer's model for activity coefficients and is validated by comparison with a variety of thermodynamic data. The model assesses the supersaturation of 13 tanks at the Savannah River Site (SRS), indicating that small amounts of gibbsite and or aluminosilicate may form. The model is also used to evaluate proposed feed streams to the Salt Waste Processing Facility for 13 years of operation. Results indicate that dilutions using 3-4 M NaOH (about 0.3-0.4 L caustic per kg feed solution) should avoid precipitation and reduce the Na{sup +} ion concentration to 5.6 M.

  15. Corrosion aspects of high-level waste disposal in salt domes

    International Nuclear Information System (INIS)

    Roerbo, K.

    1979-12-01

    In the ELSAM/ELKRAT waste management project it is planned that the high-level waste is glassified, encapsuled in canisters and finally deposited in a deep hole drilled in a salt dome. In the present report corrosion aspects of the canisters after deposition are discussed. The chemical environment will probably be a limited amount of brine coming from brine inclusions in the surrounding salt and moving up against the temperature gradient, the temperature at the canister surface being in the range of 100-150degC. The possible types of corrosion and the expected corrosion rates for a number of potential canister materials (mild steel, austenitic and ferritic stainless steels, Ni-base alloys, copper, titanium and a few combinations of materials) are discussed. Mild steel (possibly combined with an inner layer of copper or titanium) might possibly be an appropriate choice of material for the canister. (author)

  16. Assessment of Options for the Treatment of Nitrate Salt Wastes at Los Alamos National Laboratory

    International Nuclear Information System (INIS)

    Robinson, Bruce Alan; Funk, David John; Stevens, Patrice Ann

    2016-01-01

    This paper summarizes the methodology used to evaluate options for treatment of the remediated nitrate salt waste containers at Los Alamos National Laboratory. The method selected must enable treatment of the waste drums, which consist of a mixture of complex nitrate salts (oxidizer) improperly mixed with sWheat Scoop®1, an organic kitty litter and absorbent (fuel), in a manner that renders the waste safe, meets the specifications of waste acceptance criteria, and is suitable for transport and final disposal in the Waste Isolation Pilot Plant located in Carlsbad, New Mexico. A Core Remediation Team was responsible for comprehensively reviewing the options, ensuring a robust, defensible treatment recommendation. The evaluation process consisted of two steps. First, a prescreening process was conducted to cull the list on the basis for a decision of feasibility of certain potential options with respect to the criteria. Then, the remaining potential options were evaluated and ranked against each of the criteria in a consistent methodology. Numerical scores were established by consensus of the review team. Finally, recommendations were developed based on current information and understanding of the scientific, technical, and regulatory situation. A discussion of the preferred options and documentation of the process used to reach the recommended treatment options are presented.

  17. Applicability of molten salt oxidation to the destruction of actinide-contaminated wastes

    International Nuclear Information System (INIS)

    West, M.H.; Garcia, E.; Griego, W.J.; Court, D.B.; Rodriguez, L.

    1992-01-01

    A 1989 ban on incineration in the state of New Mexico caused cessation of actinide-contaminated cheesecloth, paper, and wood incineration within the Plutonium Facility (TA-55) at Los Alamos National Laboratory. Subsequently, plastic wipes were substituted for cheesecloth in the cleaning of glovebox interiors. However, waste minimization is not achieved by these measures since the wipes are discarded as Waste Isolation Pilot Plant certifiable wastes. After the ban was instituted, thermal decomposition of cheesecloth under argon at elevated temperature was examined and found satisfactory although scale of operation and speed were inferior to incineration. In 1991, the ban on incineration was lifted in New Mexico but Alamos has not chosen to pursue renewal of incineration at the Plutonium Facility. This paper reports that Los Alamos is looking from alternatives to incineration and thermal decomposition which are compatible with molten salt processing technology, historically a strength in actinide research at the Laboratory. Also, the technology must significantly reduce the volume of the waste upon treatment, i.e. waste minimization. Molten salt oxidation (MSO) has the promise of such a technology

  18. Conditioning matrices from high level waste resulting from pyrochemical processing in fluorine salt

    International Nuclear Information System (INIS)

    Grandjean, Agnes; Advocat, Thierry; Bousquet, Nicolas; Jegou, Christophe

    2007-01-01

    Separating the actinides from the fission products through reductive extraction by aluminium in a LiF/AlF 3 medium is a process investigated for pyrometallurgical reprocessing of spent fuel. The process involves separation by reductive salt-metal extraction. After dissolving the fuel or the transmutation target in a salt bath, the noble metal fission products are first extracted by contacting them with a slightly reducing metal. After extracting the metal fission products, then the actinides are selectively separated from the remaining fission products. In this hypothesis, all the unrecoverable fission products would be conditioned as fluorides. Therefore, this process will generate first a metallic waste containing the 'reducible' fission products (Pd, Mo, Ru, Rh, Tc, etc.) and a fluorine waste containing alkali-metal, alkaline-earth and rare earth fission products. Immobilization of these wastes in classical borosilicate glasses is not feasible due to the very low solubility of noble metals, and of fluoride in these hosts. Alternative candidates have therefore been developed including silicate glass/ceramic system for fluoride fission products and metallic ones for noble metal fission products. These waste-forms were evaluated for their confinement properties like homogeneity, waste loading, volatility during the elaboration process, chemical durability, etc. using appropriate techniques. (authors)

  19. Assessment of Options for the Treatment of Nitrate Salt Wastes at Los Alamos National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Robinson, Bruce Alan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Funk, David John [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stevens, Patrice Ann [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-03-17

    This paper summarizes the methodology used to evaluate options for treatment of the remediated nitrate salt waste containers at Los Alamos National Laboratory. The method selected must enable treatment of the waste drums, which consist of a mixture of complex nitrate salts (oxidizer) improperly mixed with sWheat Scoop®1, an organic kitty litter and absorbent (fuel), in a manner that renders the waste safe, meets the specifications of waste acceptance criteria, and is suitable for transport and final disposal in the Waste Isolation Pilot Plant located in Carlsbad, New Mexico. A Core Remediation Team was responsible for comprehensively reviewing the options, ensuring a robust, defensible treatment recommendation. The evaluation process consisted of two steps. First, a prescreening process was conducted to cull the list on the basis for a decision of feasibility of certain potential options with respect to the criteria. Then, the remaining potential options were evaluated and ranked against each of the criteria in a consistent methodology. Numerical scores were established by consensus of the review team. Finally, recommendations were developed based on current information and understanding of the scientific, technical, and regulatory situation. A discussion of the preferred options and documentation of the process used to reach the recommended treatment options are presented.

  20. Performance Assessment of a Generic Repository in Bedded Salt for DOE-Managed Nuclear Waste

    Science.gov (United States)

    Stein, E. R.; Sevougian, S. D.; Hammond, G. E.; Frederick, J. M.; Mariner, P. E.

    2016-12-01

    A mined repository in salt is one of the concepts under consideration for disposal of DOE-managed defense-related spent nuclear fuel (SNF) and high level waste (HLW). Bedded salt is a favorable medium for disposal of nuclear waste due to its low permeability, high thermal conductivity, and ability to self-heal. Sandia's Generic Disposal System Analysis framework is used to assess the ability of a generic repository in bedded salt to isolate radionuclides from the biosphere. The performance assessment considers multiple waste types of varying thermal load and radionuclide inventory, the engineered barrier system comprising the waste packages, backfill, and emplacement drifts, and the natural barrier system formed by a bedded salt deposit and the overlying sedimentary sequence (including an aquifer). The model simulates disposal of nearly the entire inventory of DOE-managed, defense-related SNF (excluding Naval SNF) and HLW in a half-symmetry domain containing approximately 6 million grid cells. Grid refinement captures the detail of 25,200 individual waste packages in 180 disposal panels, associated access halls, and 4 shafts connecting the land surface to the repository. Equations describing coupled heat and fluid flow and reactive transport are solved numerically with PFLOTRAN, a massively parallel flow and transport code. Simulated processes include heat conduction and convection, waste package failure, waste form dissolution, radioactive decay and ingrowth, sorption, solubility limits, advection, dispersion, and diffusion. Simulations are run to 1 million years, and radionuclide concentrations are observed within an aquifer at a point approximately 4 kilometers downgradient of the repository. The software package DAKOTA is used to sample likely ranges of input parameters including waste form dissolution rates and properties of engineered and natural materials in order to quantify uncertainty in predicted concentrations and sensitivity to input parameters. Sandia

  1. User's manual and guide to SALT3 and SALT4: two-dimensional computer codes for analysis of test-scale underground excavations for the disposal of radioactive waste in bedded salt deposits

    International Nuclear Information System (INIS)

    Lindner, E.N.; St John, C.M.; Hart, R.D.

    1984-02-01

    SALT3 and SALT4 are two-dimensional analytical/displacement-discontinuity codes designed to evaluate temperatures, deformation, and stresses associated with underground disposal of radioactive waste in bedded salt. These codes were developed by the University of Minnesota for the Office of Nuclear Waste Isolation in 1979. The present documentation describes the mathematical equations of the physical system being modeled, the numerical techniques utilized, and the organization of these computer codes. The SALT3 and SALT4 codes can simulate: (a) viscoelastic behavior in pillars adjacent to excavations; (b) transversely isotropic elastic moduli such as those exhibited by bedded or stratified rock; and (c) excavation sequence. Major advantages of these codes are: (a) computational efficiency; (b) the small amount of input data required; and (c) a creep law based on laboratory experimental data for salt. The main disadvantage is that some of the assumptions in the formulation of the codes, i.e., the homogeneous elastic half-space and temperature-independent material properties, render it unsuitable for canister-scale analysis or analysis of lateral deformation of the pillars. The SALT3 and SALT4 codes can be used for parameter sensitivity analyses of two-dimensional, repository-scale, thermomechanical response in bedded salt during the excavation, operational, and post-closure phases. It is especially useful in evaluating alternative patterns and sequences of excavation or waste canister placement. SALT3 is a refinement of an earlier code, SALT, and includes a fully anelastic creep model and thermal stress routine. SALT4 is a later version, and incorporates a revised creep model which is strain-hardening

  2. Technical support for GEIS: radioactive waste isolation in geologic formations. Volume 16. Repository preconceptual design studies: BPNL waste forms in salt

    International Nuclear Information System (INIS)

    1978-04-01

    This volume, Volume 16, ''Repository Preconceptual Design Studies: BPNL Waste Forms in Salt,'' is one of a 23 volume series, ''Technical Support for GEIS: Radioactive Waste Isolation in Geologic Formations,'' Y/OWI/TM-36, which supplements the ''Contribution to Draft Generic Environmental Impact Statement on Commercial Waste Management: Radioactive Waste Isolation in Geologic Formations,'' Y/OWI/TM-44. The series provide a more complete technical basis for the preconceptual designs, resource requirements, and environmental source terms associated with isolating commercial LWR wastes in underground repositories in salt, granite, shale and basalt. Wastes are considered from three fuel cycles: uranium and plutonium recycling, no recycling of spent fuel and uranium-only recycling. This document describes a preconceptual design for a nuclear waste storage facility in salt. The waste forms assumed to arrive at the repository were supplied by Battelle Pacific Northwest Laboratories (BPNL). The facility design consists of several chambers excavated deep within a geologic formation together with access shafts and supportive surface structures. The facility design provides for: receiving and unloading waste containers; lowering them down shafts to the mine level; transporting them to the proper storage area and emplacing them in mined storage rooms. Drawings of the facility design are contained in TM-36/17, ''Drawings for Repository Preconceptual Design Studies: BPNL Waste Forms in Salt.''

  3. Release consequence analysis for a hypothetical geologic radioactive waste repository in salt

    International Nuclear Information System (INIS)

    1979-08-01

    One subtask conducted under the INFCE program is to evaluate and compare the health and safety impacts of different fuel cycles in which all radioactive wastes (except those from mining and milling) are placed in a geologic repository in salt. To achieve this objective, INFCE Working Group 7 examined the radiologic dose to humans from geologic repositories containing waste arisings as defined for seven reference fuel cycles. This report examines the release consequences for a generic waste repository in bedded salt. The top of the salt formation and the top of the repository are assumed to be 250 and 600 m, respectively, below the surface. The hydrogeologic structure above the salt consists of two aquifers and two aquitards. The aquifers connect to a river 6.2 km from the repository. The regional gradient to the river is 1 m/km in all aquifers. Hydrologic, transport, and dose models were used to model two release scenarios for each fuel cycle, one without a major disturbance and one in which a major geologic perturbation breached the repository immediately after it was sealed. The purpose of the modeling was to predict the rate of transport of radioactive contaminants from the repository through the geosphere to the biosphere, and to determine the potential dose to humans. Of the many radionuclides in the waste, only 129 I and 226 Ra arrived at the river in sufficient concentrations for a measurable dose calculation. Radionuclide concentrations in the ground water pose no threat to man because the ground water is a concentrated brine and it is diluted by a factor of 10 6 to 10 7 upon entering the river

  4. Thermomigration of fluid inclusions in rock salt. Implications for the disposal of nuclear wastes

    International Nuclear Information System (INIS)

    Noack, W.; Runge, K.

    1984-01-01

    A mathematical model has been suggested to predict the time-dependent accumulation of brine at nuclear waste packages emplaced in a rock salt repository owing to thermomigration of brine inclusions. The model is based mainly on a description of the migration rate as a function of the temperature, temperature gradient, inclusion size and gas/liquid ratio of inclusions. Other factors are treated merely as disturbing quantities with respect to the migration rate. (author)

  5. Risk assessment of nonhazardous oil-field waste disposal in salt caverns

    International Nuclear Information System (INIS)

    Elcock, D.

    1998-01-01

    In 1996, Argonne National Laboratory (ANL) conducted a preliminary technical and legal evaluation of disposing of nonhazardous oil-field wastes (NOW) into salt caverns. Argonne determined that if caverns are sited and designed well, operated carefully, closed properly, and monitored routinely, they could be suitable for disposing of oil-field wastes. On the basis of these findings, Argonne subsequently conducted a preliminary evaluation of the possibility that adverse human health effects (carcinogenic and noncarcinogenic) could result from exposure to contaminants released from the NOW disposed of in domal salt caverns. Steps used in this evaluation included the following: identifying potential contaminants of concern, determining how humans could be exposed to these contaminants, assessing contaminant toxicities, estimating contaminant intakes, and calculating human cancer and noncancer risk estimates. Five postclosure cavern release scenarios were assessed. These were inadvertent cavern intrusion, failure of the cavern seal, failure of the cavern through cracks, failure of the cavern through leaky interbeds, and a partial collapse of the cavern roof. Assuming a single, generic, salt cavern and generic oil-field wastes, potential human health effects associated with constituent hazardous substances (arsenic, benzene, cadmium, and chromium) were assessed under each of these scenarios. Preliminary results provided excess cancer risk and hazard index (referring to noncancer health effects) estimates that were well within the US Environmental Protection Agency (EPA) target range for acceptable exposure risk levels. These results led to the preliminary conclusion that from a human health perspective, salt caverns can provide an acceptable disposal method for nonhazardous oil-field wastes

  6. Risk assessment of nonhazardous oil-field waste disposal in salt caverns.

    Energy Technology Data Exchange (ETDEWEB)

    Elcock, D.

    1998-03-05

    In 1996, Argonne National Laboratory (ANL) conducted a preliminary technical and legal evaluation of disposing of nonhazardous oil-field wastes (NOW) into salt caverns. Argonne determined that if caverns are sited and designed well, operated carefully, closed properly, and monitored routinely, they could be suitable for disposing of oil-field wastes. On the basis of these findings, Argonne subsequently conducted a preliminary evaluation of the possibility that adverse human health effects (carcinogenic and noncarcinogenic) could result from exposure to contaminants released from the NOW disposed of in domal salt caverns. Steps used in this evaluation included the following: identifying potential contaminants of concern, determining how humans could be exposed to these contaminants, assessing contaminant toxicities, estimating contaminant intakes, and calculating human cancer and noncancer risk estimates. Five postclosure cavern release scenarios were assessed. These were inadvertent cavern intrusion, failure of the cavern seal, failure of the cavern through cracks, failure of the cavern through leaky interbeds, and a partial collapse of the cavern roof. Assuming a single, generic, salt cavern and generic oil-field wastes, potential human health effects associated with constituent hazardous substances (arsenic, benzene, cadmium, and chromium) were assessed under each of these scenarios. Preliminary results provided excess cancer risk and hazard index (referring to noncancer health effects) estimates that were well within the US Environmental Protection Agency (EPA) target range for acceptable exposure risk levels. These results led to the preliminary conclusion that from a human health perspective, salt caverns can provide an acceptable disposal method for nonhazardous oil-field wastes.

  7. Forecasting the space-time stability of radioactive waste isolation in salt formations

    International Nuclear Information System (INIS)

    Anderson, E.B.; Karelin, A.I.; Krivokhatsiy, A.S.; Savonenkov, V.G.

    1992-01-01

    The possibilities to use salt formations for radioactive waste isolation are realized by creating shaft-type underground repositories in these rocks in Germany and the USA. The burial safety of low- and intermediate-level wastes for several hundred years have been substantiated for the sites chosen. Specialists of different countries presented positive properties of rock salt as a medium for isolation of radionuclides. A rich experience in building subsurface structures for different purposes in salts is accumulated in our country. Detailed investigations of salt formation have shown that far from all the saliferous areas and structures may be used for constructing burial sites. One of the reasons for this limitation is a sharp difference of individual deposits by their compositions, structures, the character of deposition and the conditions of formation. The geological criteria of safety acquire special significance in connection with the necessity to isolate radionuclides having the half-loves more than 1000 years. The time intervals required for stable isolation make up millions of years and cover great cycles of the evolution of the Earth surface and biosphere

  8. Implications of thermophysical properties in geoscientific investigations for the disposal of nuclear waste in a salt dome

    International Nuclear Information System (INIS)

    Kopietz, J.

    1984-01-01

    Examples from laboratory and in-situ experiments on the thermomechanical behavior of rock salt are used to discuss the implications of thermophysical properties for disposal of nuclear waste in a salt dome. The implications of thermophysical properties are also illustrated by a brief review of geothermal investigations made within the scope of geological and hydrogeological exploration of the Gorleben salt dome in northern Germany. High-resolution temperature measurements performed in shallow and deep boreholes drilled for the exploration of the Gorleben salt dome, together with thermal conductivity measurements on representative core samples from these boreholes, are contributing to a determination of groundwater flow in the covering layers of the salt dome and to the identification of zones of impurity (eg carnallitite layers) within the salt structure. Data from these experiments are used for setting up numerical models for heat propagation around a prospective waste repository in the Gorleben salt dome. Long-term creep experiments on samples of rock salt at up to 400 deg C are used to derive constitutive relations on the creep behavior of salt. In-situ heating experiments are being conducted in the Asse salt mine to determine the effect of a heat source on the integrity of the surrounding salt rock. (author)

  9. Hydronephrosis causes salt-sensitive hypertension and impaired renal concentrating ability in mice

    DEFF Research Database (Denmark)

    Carlström, M; Sällström, J; Skøtt, O

    2007-01-01

    AIM: Hypertension is a common disease in the industrialized world and approximately 5% of all cases are secondary to kidney malfunction. We have recently shown that hydronephrosis due to partial unilateral ureteral obstruction (PUUO) causes salt-sensitive hypertension in rats. The mechanisms...... are still unclear, but appear to be intrarenal and primarily located to the diseased kidney. In the present study, we have developed a model for PUUO to study if hydronephrotic mice develop salt-sensitive hypertension. METHODS: PUUO was created in 3-week-old mice (C57bl/6J). Blood pressure and heart rate...... salt-sensitive hypertension that correlated to the degree of hydronephrosis. In hydronephrotic animals, blood pressure increased from 114 +/- 1 mmHg on normal salt diet to 120 +/- 2 mmHg on high salt diet, compared with 103 +/- 1 to 104 +/- 1 in controls. Hydronephrotic animals showed increased...

  10. An experimental study on Sodalite and SAP matrices for immobilization of spent chloride salt waste

    Science.gov (United States)

    Giacobbo, Francesca; Da Ros, Mirko; Macerata, Elena; Mariani, Mario; Giola, Marco; De Angelis, Giorgio; Capone, Mauro; Fedeli, Carlo

    2018-02-01

    In the frame of Generation IV reactors a renewed interest in pyro-processing of spent nuclear fuel is underway. Molten chloride salt waste arising from the recovering of uranium and plutonium through pyro-processing is one of the problematic wastes for direct application of vitrification or ceramization. In this work, Sodalite and SAP have been evaluated and compared as potential matrices for confinement of spent chloride salt waste coming from pyro-processing. To this aim Sodalite and SAP were synthesized both in pure form and mixed with different glass matrices, i.e. commercially available glass frit and borosilicate glass. The confining matrices were loaded with mixed chloride salts to study their retention capacities with respect to the elements of interest. The matrices were characterized and leached for contact times up to 150 days at room temperature and at 90 °C. SEM analyses were also performed in order to compare the matrix surface before and after leaching. Leaching results are discussed and compared in terms of normalized releases with similar results reported in literature. According to this comparative study the SAP matrix with glass frit binder resulted in the best matrix among the ones studied, with respect to retention capacities for both matrix and spent fuel elements.

  11. Risk analyses for disposing nonhazardous oil field wastes in salt caverns

    Energy Technology Data Exchange (ETDEWEB)

    Tomasko, D.; Elcock, D.; Veil, J.; Caudle, D.

    1997-12-01

    Salt caverns have been used for several decades to store various hydrocarbon products. In the past few years, four facilities in the US have been permitted to dispose nonhazardous oil field wastes in salt caverns. Several other disposal caverns have been permitted in Canada and Europe. This report evaluates the possibility that adverse human health effects could result from exposure to contaminants released from the caverns in domal salt formations used for nonhazardous oil field waste disposal. The evaluation assumes normal operations but considers the possibility of leaks in cavern seals and cavern walls during the post-closure phase of operation. In this assessment, several steps were followed to identify possible human health risks. At the broadest level, these steps include identifying a reasonable set of contaminants of possible concern, identifying how humans could be exposed to these contaminants, assessing the toxicities of these contaminants, estimating their intakes, and characterizing their associated human health risks. The contaminants of concern for the assessment are benzene, cadmium, arsenic, and chromium. These were selected as being components of oil field waste and having a likelihood to remain in solution for a long enough time to reach a human receptor.

  12. ICP-MS nebulizer performance for analysis of SRS high salt simulated radioactive waste tank solutions (number-sign 3053)

    International Nuclear Information System (INIS)

    Jones, V.D.

    1997-01-01

    High Level Radioactive Waste Tanks at the Savannah River Site are high in salt content. The cross-flow nebulizer provided the most stable signal for all salt matrices with the smallest signal loss/suppression due to this matrix. The DIN exhibited a serious lack of tolerance for TDS; possibly due to physical de-tuning of the nebulizer efficiency

  13. The HAW-project: Demonstration facility for the disposal of high-level waste in salt

    International Nuclear Information System (INIS)

    Rothfuchs, T.; Duijves, K.A.

    1990-04-01

    The HAW-project plants the testwise emplacement of 30 vitrified highly radioactive canisters containing Cs-137 and Sr-90 at the 800 m level of the Asse salt mine for a testing period of approximately five years. The major objective of this project is the pilot testing and demonstration of safe methods for the final disposal of high-level radioactive waste (HAW) in geological salt formations. During the years 1985 to 1989 the underground test field was excavated, the measuring equipment installed, and two preceedings inactive electrical tests taken into operation. Furthermore, the components of a system for transportation and emplacement of highly radioactive canisters was fabricated, installed, and preliminarily tested. After some delays in the licensing procedure the emplacement of the 30 radioactive canisters is now envisaged for early 1991. For handling of the radioactive canisters and their emplacement into the boreholes a system consisting of a transport cask, a transport vehicle, a disposal machine, and of a borehole slider has been developed and will be tested. The actual scientific investigation programme is based on the estimation and observation of the interaction between the radioactive canisters and the rock salt. This programme includes measurement of thermally and radiolytically induced water and gas release from the rock salt and the radiolytical decomposition of salt minerals. Also the thermally induced stress and deformation fields in the surrounding rock mass will be investigated carefully. (orig./HP)

  14. Thermal properties of fly ash substituted slag cement waste forms for disposal of Savannah River Plant salt waste

    International Nuclear Information System (INIS)

    Roy, D.M.; Kaushal, S.; Licastro, P.H.; Langton, C.A.

    1985-01-01

    Waste processing at the Savannah River Plant will involve reconstitution of the salts (NaNO 3 , NaNO 2 , NaOH, etc.) into a concentrated solution (32 weight percent salts) followed by solidification in a cement-based waste form for burial. The stability and mechanical durability of such a 'saltstone monolith' will depend largely on the temperature reached due to heat of hydration and the thermal properties of the waste form. Fly ash has been used as an inexpensive constituent and to moderate the hydration and setting processes so as to avoid reaching prohibitively high temperatures which could cause thermal stresses. Both high-calcium and low-calcium fly ashes have been studied for this purpose. Other constituents of these mixes include granulated blast furnace slag and finely crushed limestone. Adiabatic temperature increase and thermal conductivity of these mixes have been studied and related x-ray diffraction and scanning electron microscopy studies carried out to understand the hydration process

  15. Defense Waste Processing Facility (DWPF), Modular CSSX Unit (CSSX), and Waste Transfer Line System of Salt Processing Program (U)

    International Nuclear Information System (INIS)

    CHANG, ROBERT

    2006-01-01

    All of the waste streams from ARP, MCU, and SWPF processes will be sent to DWPF for vitrification. The impact these new waste streams will have on DWPF's ability to meet its canister production goal and its ability to support the Salt Processing Program (ARP, MCU, and SWPF) throughput needed to be evaluated. DWPF Engineering and Operations requested OBU Systems Engineering to evaluate DWPF operations and determine how the process could be optimized. The ultimate goal will be to evaluate all of the Liquid Radioactive Waste (LRW) System by developing process modules to cover all facilities/projects which are relevant to the LRW Program and to link the modules together to: (1) study the interfaces issues, (2) identify bottlenecks, and (3) determine the most cost effective way to eliminate them. The results from the evaluation can be used to assist DWPF in identifying improvement opportunities, to assist CBU in LRW strategic planning/tank space management, and to determine the project completion date for the Salt Processing Program

  16. Results Of The Extraction-Scrub-Strip Testing Using An Improved Solvent Formulation And Salt Waste Processing Facility Simulated Waste

    International Nuclear Information System (INIS)

    Peters, T.; Washington, A.; Fink, S.

    2012-01-01

    The Office of Waste Processing, within the Office of Technology Innovation and Development, is funding the development of an enhanced solvent - also known as the next generation solvent (NGS) - for deployment at the Savannah River Site to remove cesium from High Level Waste. The technical effort is a collaborative effort between Oak Ridge National Laboratory (ORNL) and Savannah River National Laboratory (SRNL). As part of the program, the Savannah River National Laboratory (SRNL) has performed a number of Extraction-Scrub-Strip (ESS) tests. These batch contact tests serve as first indicators of the cesium mass transfer solvent performance with actual or simulated waste. The test detailed in this report used simulated Tank 49H material, with the addition of extra potassium. The potassium was added at 1677 mg/L, the maximum projected (i.e., a worst case feed scenario) value for the Salt Waste Processing Facility (SWPF). The results of the test gave favorable results given that the potassium concentration was elevated (1677 mg/L compared to the current 513 mg/L). The cesium distribution value, DCs, for extraction was 57.1. As a comparison, a typical D Cs in an ESS test, using the baseline solvent formulation and the typical waste feed, is ∼15. The Modular Caustic Side Solvent Extraction Unit (MCU) uses the Caustic-Side Solvent Extraction (CSSX) process to remove cesium (Cs) from alkaline waste. This process involves the use of an organic extractant, BoBCalixC6, in an organic matrix to selectively remove cesium from the caustic waste. The organic solvent mixture flows counter-current to the caustic aqueous waste stream within centrifugal contactors. After extracting the cesium, the loaded solvent is stripped of cesium by contact with dilute nitric acid and the cesium concentrate is transferred to the Defense Waste Processing Facility (DWPF), while the organic solvent is cleaned and recycled for further use. The Salt Waste Processing Facility (SWPF), under

  17. X-linked lissencephaly with abnormal genitalia associated with renal phosphate wasting.

    Science.gov (United States)

    Hahn, A; Gross, C; Uyanik, G; Hehr, U; Hügens-Penzel, M; Alzen, G; Neubauer, B A

    2004-06-01

    X-linked lissencephaly with abnormal genitalia (XLAG) is a rare disorder caused by mutations in the aristaless-related homeobox (ARX) gene. We report on the clinical data of a boy with a 1-bp deletion (790 delC) resulting in a frame shift in the ARX gene and prolonged survival until age 18 months. Similar to other patients, the boy showed postnatal microcephaly, hypothalamic dysfunction, intractable neonatal seizures, and chronic diarrhoea. In addition, he suffered from exocrine pancreatic insufficiency and renal phosphate wasting became apparent from age 5 months, both of which have not been described previously in XLAG. This allows us to speculate that the phenotype of XLAG is more complex than hitherto known and may include renal phosphate wasting which might not have been observed in other patients due to early death.

  18. The function of packing materials in a high-level nuclear waste repository and some candidate materials: Salt Repository Project

    International Nuclear Information System (INIS)

    Bunnell, L.R.; Shade, J.W.

    1987-03-01

    Packing materials should be included in waste package design for a high-level nuclear waste repository in salt. A packing material barrier would increase confidence in the waste package by alleviating possible shortcomings in the present design and prolonging confinement capabilities. Packing materials have been studied for uses in other geologic repositories; appropriately chosen, they would enhance the confinement capabilities of salt repository waste packages in several ways. Benefits of packing materials include retarding or chemically modifying brines to reduce corrosion of the waste package, providing good thermal conductivity between the waste package and host rock, retarding or absorbing radionuclides, and reducing the massiveness of the waste package. These benefits are available at low percentage of total repository cost, if the packing material is properly chosen and used. Several candidate materials are being considered, including oxides, hydroxides, silicates, cement-based mixtures, and clay mixtures. 18 refs

  19. Differential regulation of renal prostaglandin receptor mRNAs by dietary salt intake in the rat

    DEFF Research Database (Denmark)

    Jensen, B L; Mann, Birgitte; Skøtt, O

    1999-01-01

    and cells by ribonuclease protection assay and reverse transcription-polymerase chain reaction analysis. Functional correlates were studied by measurement of PGE2-induced cAMP formation and renin secretion in juxtaglomerular (JG) cells isolated from animals on various salt intakes. RESULTS: EP1 and EP3......BACKGROUND: In this study, we tested the hypothesis that prostaglandin (PG) receptor expression in the rat kidney is subject to physiological regulation by dietary salt intake. METHODS: Rats were fed diets with 0.02 or 4% NaCl for two weeks. PG receptor expression was assayed in kidney regions...... did not affect the expression of EP1 or IP receptors, whereas EP4 transcripts in glomeruli were increased twofold by salt deprivation. Consistent with this, we found that PGE2-evoked cAMP production and renin secretion by JG cells from salt-deprived animals were significantly higher compared...

  20. UK-Nuclear decommissioning authority and US Salt-stone waste management issues

    International Nuclear Information System (INIS)

    Lawless, William; Whitton, John

    2007-01-01

    Available in abstract form only. Full text of publication follows: We update two case studies of stakeholder issues in the UK and US. Earlier versions were reported at Waste Management 2006 and 2007 and at ICEM 2005. UK: The UK nuclear industry has begun to consult stakeholders more widely in recent years. Historically, methods of engagement within the industry have varied, however, recent discussions have generally been carried out with the explicit understanding that engagement with stakeholders will be 'dialogue based' and will 'inform' the final decision made by the decision maker. Engagement is currently being carried out at several levels within the industry; at the national level (via the Nuclear Decommissioning Authority's (NDA) National Stakeholder Group (NSG)); at a local site level (via Site Stakeholder Groups) and at a project level (usually via the Best Practicable Environmental Option process (BPEO)). This paper updates earlier results by the co-author with findings from a second questionnaire issued to the NSG in Phase 2 of the engagement process. An assessment is made regarding the development of stakeholder perceptions since Phase 1 towards the NDA process. US: The US case study reviews the resolution of issues on salt-stone by Department of Energy's (DOE) Savannah River Site (SRS) Citizens Advisory Board (CAB), in Aiken, SC. Recently, SRS-CAB encouraged DOE and South Carolina's regulatory Department of Health and Environmental Control (SC-DHEC) to resolve a conflict preventing SC-DHEC from releasing a draft permit to allow SRS to restart salt-stone operations. It arose with a letter sent from DOE blaming the Governor of South Carolina for delay in restarting salt processing. In reply, the Governor blamed DOE for failing to assure that Salt Waste Processing Facility (SWPF) would be built. SWPF is designed to remove most of the radioactivity from HLW prior to vitrification, the remaining fraction destined for salt-stone. (authors)

  1. NRC Monitoring of Salt Waste Disposal at the Savannah River Site - 13147

    Energy Technology Data Exchange (ETDEWEB)

    Pinkston, Karen E.; Ridge, A. Christianne; Alexander, George W.; Barr, Cynthia S.; Devaser, Nishka J.; Felsher, Harry D. [U.S. Nuclear Regulatory Commission (United States)

    2013-07-01

    As part of monitoring required under Section 3116 of the Ronald W. Reagan National Defense Authorization Act for Fiscal Year 2005 (NDAA), the NRC staff reviewed an updated DOE performance assessment (PA) for salt waste disposal at the Saltstone Disposal Facility (SDF). The NRC staff concluded that it has reasonable assurance that waste disposal at the SDF meets the 10 CFR 61 performance objectives for protection of individuals against intrusion (chap.61.42), protection of individuals during operations (chap.61.43), and site stability (chap.61.44). However, based on its evaluation of DOE's results and independent sensitivity analyses conducted with DOE's models, the NRC staff concluded that it did not have reasonable assurance that DOE's disposal activities at the SDF meet the performance objective for protection of the general population from releases of radioactivity (chap.61.41) evaluated at a dose limit of 0.25 mSv/yr (25 mrem/yr) total effective dose equivalent (TEDE). NRC staff also concluded that the potential dose to a member of the public is expected to be limited (i.e., is expected to be similar to or less than the public dose limit in chap.20.1301 of 1 mSv/yr [100 mrem/yr] TEDE) and is expected to occur many years after site closure. The NRC staff used risk insights gained from review of the SDF PA, its experience monitoring DOE disposal actions at the SDF over the last 5 years, as well as independent analysis and modeling to identify factors that are important to assessing whether DOE's disposal actions meet the performance objectives. Many of these factors are similar to factors identified in the NRC staff's 2005 review of salt waste disposal at the SDF. Key areas of interest continue to be waste form and disposal unit degradation, the effectiveness of infiltration and erosion controls, and estimation of the radiological inventory. Based on these factors, NRC is revising its plan for monitoring salt waste disposal at the SDF in

  2. NRC Monitoring of Salt Waste Disposal at the Savannah River Site - 13147

    International Nuclear Information System (INIS)

    Pinkston, Karen E.; Ridge, A. Christianne; Alexander, George W.; Barr, Cynthia S.; Devaser, Nishka J.; Felsher, Harry D.

    2013-01-01

    As part of monitoring required under Section 3116 of the Ronald W. Reagan National Defense Authorization Act for Fiscal Year 2005 (NDAA), the NRC staff reviewed an updated DOE performance assessment (PA) for salt waste disposal at the Saltstone Disposal Facility (SDF). The NRC staff concluded that it has reasonable assurance that waste disposal at the SDF meets the 10 CFR 61 performance objectives for protection of individuals against intrusion (chap.61.42), protection of individuals during operations (chap.61.43), and site stability (chap.61.44). However, based on its evaluation of DOE's results and independent sensitivity analyses conducted with DOE's models, the NRC staff concluded that it did not have reasonable assurance that DOE's disposal activities at the SDF meet the performance objective for protection of the general population from releases of radioactivity (chap.61.41) evaluated at a dose limit of 0.25 mSv/yr (25 mrem/yr) total effective dose equivalent (TEDE). NRC staff also concluded that the potential dose to a member of the public is expected to be limited (i.e., is expected to be similar to or less than the public dose limit in chap.20.1301 of 1 mSv/yr [100 mrem/yr] TEDE) and is expected to occur many years after site closure. The NRC staff used risk insights gained from review of the SDF PA, its experience monitoring DOE disposal actions at the SDF over the last 5 years, as well as independent analysis and modeling to identify factors that are important to assessing whether DOE's disposal actions meet the performance objectives. Many of these factors are similar to factors identified in the NRC staff's 2005 review of salt waste disposal at the SDF. Key areas of interest continue to be waste form and disposal unit degradation, the effectiveness of infiltration and erosion controls, and estimation of the radiological inventory. Based on these factors, NRC is revising its plan for monitoring salt waste disposal at the SDF in coordination with South

  3. Buckling design criteria for waste package disposal containers in mined salt repositories: Technical report

    International Nuclear Information System (INIS)

    Mallett, R.H.

    1986-12-01

    This report documents analytical and experimental results from a survey of the technical literature on buckling of thick-walled cylinders under external pressure. Based upon these results, a load factor is suggested for the design of waste package containers for disposal of high-level radioactive waste in repositories mined in salt formations. The load factor is defined as a ratio of buckling pressure to allowable pressure. Specifically, a load factor which ranges from 1.5 for plastic buckling to 3.0 for elastic buckling is included in a set of proposed buckling design criteria for waste disposal containers. Formulas are given for buckling design under axisymmetric conditions. Guidelines are given for detailed inelastic buckling analyses which are generally required for design of disposal containers

  4. Analytical Chemistry and Materials Characterization Results for Debris Recovered from Nitrate Salt Waste Drum S855793

    Energy Technology Data Exchange (ETDEWEB)

    Martinez, Patrick Thomas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Chamberlin, Rebecca M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Schwartz, Daniel S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Worley, Christopher Gordon [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Garduno, Katherine [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Lujan, Elmer J. W. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Borrego, Andres Patricio [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Castro, Alonso [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Colletti, Lisa Michelle [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Fulwyler, James Brent [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Holland, Charlotte S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Keller, Russell C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Klundt, Dylan James [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Martinez, Alexander [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Martin, Frances Louise [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Montoya, Dennis Patrick [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Myers, Steven Charles [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Porterfield, Donivan R. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Schake, Ann Rene [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Schappert, Michael Francis [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Soderberg, Constance B. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Spencer, Khalil J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stanley, Floyd E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Thomas, Mariam R. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Townsend, Lisa Ellen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Xu, Ning [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-09-16

    Solid debris was recovered from the previously-emptied nitrate salt waste drum S855793. The bulk sample was nondestructively assayed for radionuclides in its as-received condition. Three monoliths were selected for further characterization. Two of the monoliths, designated Specimen 1 and 3, consisted primarily of sodium nitrate and lead nitrate, with smaller amounts of lead nitrate oxalate and lead oxide by powder x-ray diffraction. The third monolith, Specimen 2, had a complex composition; lead carbonate was identified as the predominant component, and smaller amounts of nitrate, nitrite and carbonate salts of lead, magnesium and sodium were also identified. Microfocused x-ray fluorescence (MXRF) mapping showed that lead was ubiquitous throughout the cross-sections of Specimens 1 and 2, while heteroelements such as potassium, calcium, chromium, iron, and nickel were found in localized deposits. MXRF examination and destructive analysis of fragments of Specimen 3 showed elevated concentrations of iron, which were broadly distributed through the sample. With the exception of its high iron content and low carbon content, the chemical composition of Specimen 3 was within the ranges of values previously observed in four other nitrate salt samples recovered from emptied waste drums.

  5. Conceptual waste package interim product specifications and data requirements for disposal of glass commercial high-level waste forms in salt geologic repositories

    International Nuclear Information System (INIS)

    1983-10-01

    The conceptual waste package interim product specifications and data requirements presented are applicable to the reference glass composition described in PNL-3838 and carbon steel canister described in ONWI-438. They provide preliminary numerical values for the commercial high-level waste form parameters and properties identified in the waste form performance specification for geologic isolation in salt repositories. Subject areas treated include containment and isolation, operational period safety, criticality control, waste form/production canister identification, and waste package performance testing requirements. This document was generated for use in the development of conceptual waste package designs in salt. It will be revised as additional data, analyses and regulatory requirements become available. 13 references, 1 figure

  6. Bibliography of studies for the Salt Repository Project Office of the Civilian Radioactive Waste Management Program, April 1978-May 1986

    International Nuclear Information System (INIS)

    1986-10-01

    DOE/CH/10140-05 is an annotated bibliography of approved reports that have been produced for the US Department of Energy Salt Repository Project Office of the Civilian Radioactive Waste Management Program since April 1978. This document is intended for use by the US Department of Energy, State and local officials, the US Nuclear Regulatory Commission, contractors to the Office of Nuclear Waste Isolation, concerned citizens, and others who need a comprehensive listing of reports related to a nuclear waste repository in salt. This document consists of a main report listing, appendixes with Work Breakdown Structure lists, and a topical index

  7. Bibliography of studies for the Salt Repository Project Office of the Civilian Radioactive Waste Management Program, April 1978-December 1986

    International Nuclear Information System (INIS)

    1987-06-01

    This document is an annotated bibliography of approved reports that have been produced for the US Department of Energy Salt Repository Project Office of the Civilian Radioactive Waste Management Program since April 1978. This document is intended for use by the US Department of Energy, State and local officials, the US Nuclear Regulatory Commission, contractors to the Office of Nuclear Waste Isolation, concerned citizens, and others who need a comprehensive listing of reports related to a nuclear waste repository in salt. This document consists of a main report listing, appendixes with Work Breakdown Structure lists, and a topical index

  8. Comparison of the salt domes Asse and Gorleben with regard to their suitability for the final storage of radoactive wastes

    International Nuclear Information System (INIS)

    Deisenroth, Norbert; Kokorsch, Rudolf

    2012-01-01

    In Germany, the search for a proper solution to the issue of final disposal of radioactive wastes is complicated by political leaders. The Gorleben moratorium from October 2000 delayed the proper solution unnecessary to ten years. Asse proves that salt domes such as Gorleben do not offer a permanent partitioning of the waste over the biosphere. With this in mind, the authors of the contribution under consideration compare the two salt domes Gorleben and Asse from a mining and geological point of view based on publicly available data with regard to their suitability for the disposal of radioactive waste.

  9. Hydrostatic and shear consolidation tests with permeability measurements on Waste Isolation Pilot Plant crushed salt

    International Nuclear Information System (INIS)

    Brodsky, N.S.

    1994-03-01

    Crushed natural rock salt is a primary candidate for use as backfill and barrier material at the Waste Isolation Pilot Plant (WIPP) and therefore Sandia National Laboratories (SNL) has been pursuing a laboratory program designed to quantify its consolidation properties and permeability. Variables that influence consolidation rate that have been examined include stress state and moisture content. The experimental results presented in this report complement existing studies and work in progress conducted by SNL. The experiments described in this report were designed to (1) measure permeabilities of consolidated specimens of crushed salt, (2) determine the influence of brine saturation on consolidation under hydrostatic loads, and 3) measure the effects of small applied shear stresses on consolidation properties. The laboratory effort consisted of 18 individual tests: three permeability tests conducted on specimens that had been consolidated at Sandia, six hydrostatic consolidation and permeability tests conducted on specimens of brine-saturated crushed WIPP salt, and nine shear consolidation and permeability tests performed on crushed WIPP salt specimens containing 3 percent brine by weight. For hydrostatic consolidation tests, pressures ranged from 1.72 MPa to 6.90 MPa. For the shear consolidation tests, confining pressures were between 3.45 MPa and 6.90 MPa and applied axial stress differences were between 0.69 and 4.14 MPa. All tests were run under drained conditions at 25 degrees C

  10. Compatibility tests between Solar Salt and thermal storage ceramics from inorganic industrial wastes

    International Nuclear Information System (INIS)

    Motte, Fabrice; Falcoz, Quentin; Veron, Emmanuel; Py, Xavier

    2015-01-01

    Highlights: • ESEM and XRD characterizations have been performed. • Compatibility of these ceramics with the conventional binary Solar Salt is tested at 500 °C. • Tested ceramics have relevant properties to store thermal energy up to 1000 °C. • Feasibility of using ceramics as filler materials in thermocline is demonstrated. - Abstract: This paper demonstrates the feasibility of using several post-industrial ceramics as filler materials in a direct thermocline storage configuration. The tested ceramics, coming from several industrial processes (asbestos containing waste treatment, coal fired power plants or metallurgic furnaces) demonstrate relevant properties to store thermal energy by sensible heat up to 1000 °C. Thus, they represent at low-cost a promising, efficient and sustainable approach for thermal energy storage. In the present study, the thermo-chemical compatibility of these ceramics with the conventional binary Solar Salt is tested at medium temperature (500 °C) under steady state. In order to determine the feasibility of using such ceramics as filler material, Environmental Scanning Electron Microscopy (ESEM) and X-Ray Diffraction (XRD) characterizations have been performed to check for their chemical and structural evolution during corrosion tests. The final objective is to develop a molten salt thermocline direct storage system using low-cost shaped ceramic as structured filler material. Most of the tested ceramics present an excellent corrosion resistance in molten Solar Salt and should significantly decrease the current cost of concentrated solar thermal energy storage system

  11. Preliminary investigation results as applied to utilization of Ukrainian salt formations for disposal of high-level radioactive waste

    International Nuclear Information System (INIS)

    Shekhunova, S.B.; Khrushchov, D.P.; Petrichenko, O.I.

    1994-01-01

    The salt-bearing formations have been investigated in five regions of Ukraine. Upper Devonian and Lower Permian evaporite formations in Dnieper-Donets Depression and in the NW part of Donets basin are considered to be promising for disposal of high-level radioactive waste (HLRW). Rock salt occurs there either as bedded salts or as salt pillows and salt diapirs. Preliminary studies have resulted in selection of several candidate sites that show promise for construction of a subsurface pilot lab. Ten salt domes and two sites in bedded salts have been proposed for further exploration. Based on microstructural studies it is possible to separate the body of a salt structure and to locate within its limits the rock salt structure and to locate within its limits the rock salt blocks of different genesis, i.e.: (a) blocks characteristic of initial undisturbed sedimentary structure; (b) flow zones; (c) sliding planes; (d) bodies of loose or uncompacted rock salt. Ultramicrochemical examination of inclusions in halite have shown that they are composed of more than 40 minerals. It is emphasized that to assess suitability of a structure for construction of subsurface lab, and also the potential construction depth intervals, account should be taken of the results of ultra microchemical and microstructural data

  12. Permian salt dissolution, alkaline lake basins, and nuclear-waste storage, Southern High Plains, Texas and New Mexico

    International Nuclear Information System (INIS)

    Reeves, C.C. Jr.; Temple, J.M.

    1986-01-01

    Areas of Permian salt dissolution associated with 15 large alkaline lake basins on and adjacent to the Southern High Plains of west Texas and eastern New Mexico suggest formation of the basins by collapse of strata over the dissolution cavities. However, data from 6 other alkaline basins reveal no evidence of underlying salt dissolution. Thus, whether the basins were initiated by subsidence over the salt dissolution areas or whether the salt dissolution was caused by infiltration of overlying lake water is conjectural. However, the fact that the lacustrine fill in Mound Lake greatly exceeds the amount of salt dissolution and subsidence of overlying beds indicates that at least Mound Lake basin was antecedent to the salt dissolution. The association of topography, structure, and dissolution in areas well removed from zones of shallow burial emphasizes the susceptibility of Permian salt-bed dissolution throughout the west Texas-eastern New Mexico area. Such evidence, combined with previous studies documenting salt-bed dissolution in areas surrounding a proposed high-level nuclear-waste repository site in Deaf Smith County, Texas, leads to serious questions about the rationale of using salt beds for nuclear-waste storage

  13. Tank Waste Transport Stability: Summaries of Hanford Slurry and Salt-Solution Studies in FY 2000

    Energy Technology Data Exchange (ETDEWEB)

    Welch, T.D.

    2002-07-08

    This report is a collection of summary articles on FY 2000 studies of slurry transport and salt-well pumping related to Hanford tank waste transfers. These studies are concerned with the stability (steady, uninterrupted flow) of tank waste transfers, a subset of the Department of Energy (DOE) Tanks Focus Area Tank (TFA) Waste Chemistry effort. This work is a collaborative effort of AEA Technology plc, the Diagnostic Instrumentation and Analysis Laboratory at Mississippi State University (DIAL-MSU), the Hemispheric Center for Environmental Technology at Florida International University (HCET-FIU), Numatec Hanford Corporation (NHC), and the Oak Ridge National Laboratory (ORNL). The purpose of this report is to provide, in a single document, an overview of these studies to help the reader identify contacts and resources for obtaining more detailed information and to help promote useful interchanges between researchers and users. Despite over 50 years of experience in transporting radioactive tank wastes to and from equipment and tanks at the Department of Energy's Hanford, Savannah River, and Oak Ridge sites, waste slurry transfer pipelines and process piping become plugged on occasion. At Hanford, several tank farm pipelines are no longer in service because of plugs. At Savannah River, solid deposits in the outlet line of the 2H evaporator have resulted in an unplanned extended downtime. Although waste transfer criteria and guidelines intended to prevent pipeline plugging are in place, they are not always adequate. To avoid pipeline plugging in the future, other factors that are not currently embodied in the transfer criteria may need to be considered. The work summarized here is being conducted to develop a better understanding of the chemical and waste flow dynamics during waste transfer. The goal is to eliminate pipeline plugs by improving analysis and engineering tools in the field that incorporate this understanding.

  14. The HAW project: demonstration facility for the disposal of high-level waste in salt

    International Nuclear Information System (INIS)

    Rothfuchs, T.

    1991-01-01

    This report is the so-called Synthesis report 1985-1989 of the international HAW project performed in the 800 m level of the ASSE salt mine in the Federal Republic of Germany. The major objective of this project is the pilot testing and demonstration of safe methods for the final disposal of high-level radioactive waste in geological salt-deposits. The HAW-project is carried out by the GSF-Institut fuer Tieflagerung (IFT) in cooperation with the French Agence Nationale pour la Gestion des Dechets Radioactifs (ANDRA); the Spanish Empresa Nacional de Residuos Radioactivos S.A (ENRESA) and the Netherlands Energy Research Foundation (ECN). During the years 1985 to 1989 the underground test field was excavated and after some delays in the licensing procedure, the emplacement of 30 vitrified highly radioactive canisters (containers) is now envisaged for early 1991. 32 refs; 76 figs., 11 tabs

  15. Comparison of temperature calculations for an arbitrary high-level waste disposal configuration in salt formations

    International Nuclear Information System (INIS)

    Kevenaar, J.W.A.M.; Janssen, L.G.J.; Ploumen, P.; Winske, P.

    1979-05-01

    The objective of this report is the comparison of the results of temperature analyses for an arbitrary high-level radioactive waste disposal configuration in salt formations. The analyses were carried out at the RWTH and ECN. The computer programs used are based on finite difference and finite element techniques. From the local temperature analyses that were intended to check the solution techniques, it could be concluded that both finite difference and finite elements are capable to analyse this type of problems. From the global temperature analyses it could be concluded that both analysis approaches: temperature dependent and iteratively determined temperature independent material properties, are suited to analyse the global temperature distribution in the salt formation

  16. The HAW project: demonstration facility for the disposal of high-level waste in salt

    International Nuclear Information System (INIS)

    Rothfuchs, T.

    1991-01-01

    This publication is the interim report 1988-89 of the international HAW project performed in the 800 m level of the Asse salt mine in the Federal Republic of Germany. The major objective of this project is the pilot testing and demonstration of safe methods for the final disposal of high-level radioactive waste in geological salt deposits. The HAW-project is carried out by the GSF-Institut fuer Tieflagerung (IFT) in cooperation with the French Agence Nationale pour la Gestion des Dechets Radioactifs (ANDRA); the Spanish Empresa Nacional de Residuos Radiactivos S.A. (ENRESA) and the Netherlands Energy Research Foundation (ECN). After some delays in the licensing procedure the emplacement of 30 vitrified highly radioactive canisters (containers) is now envisaged for early 1991. 20 refs.; 92 figs.; 14 tabs

  17. Leaching due to hygroscopic water uptake in cemented waste containing soluble salts

    DEFF Research Database (Denmark)

    Brodersen, K.

    1992-01-01

    conditions, condensation of water vapour will result in generation of a certain amount of liquid in the form of a strong salt solution. The volume of liquid may well exceed the storage capacity of the pore system in the cemented material and in the release of a limited amount of free contaminated solution......Considerable amounts of easily soluble salts such as sodium nitrate, sulphate, or carbonate are introduced into certain types of cemented waste. When such materials are stored in atmospheres with high relative humidity or disposed or by shallow land burial under unsaturated, but still humid....... A model of the quantitative aspects for the equilibrium situation is presented. Experiments with hygroscopic water uptake support the model and give indications about the rate of the process. The release mechanism is only thought to be important for radionuclides which are not fixed in a low...

  18. Development and characterization of new high-level waste form containing LiCl KCl eutectic salts for achieving waste minimization from pyroprocessing

    International Nuclear Information System (INIS)

    Cho, Yong Zun; Kim, In Tae; Park, Hwan Seo; Ahn, Byeung Gil; Eun, Hee Chul; Son, Seock Mo; Ah, Su Na

    2011-12-01

    The purpose of this project is to develop new high level waste (HLW) forms and fabrication processes to dispose of active metal fission products that are removed from electrorefiner salts in the pyroprocessing based fuel cycle. The current technology for disposing of active metal fission products in pyroprocessing involves non selectively discarding of fission product loaded salt in a glass-bonded sodalite ceramic waste form. Selective removal of fission products from the molten salt would greatly minimize the amount of HLW generated and methods were developed to achieve selective separation of fission products during a previous I NERI research project (I NERI 2006 002 K). This I NERI project proceeds from the previous project with the development of suitable waste forms to immobilize the separated fission products. The Korea Atomic Energy Research Institute (KAERI) has focused primarily on developing these waste forms using surrogate waste materials, while the Idaho National Laboratory (INL) has demonstrated fabrication of these waste forms using radioactive electrorefiner salts in hot cell facilities available at INL. Testing and characterization of these radioactive materials was also performed to determine the physical, chemical, and durability properties of the waste forms

  19. Technical support for GEIS: radioactive waste isolation in geologic formations. Volume 8. Repository preconceptual design studies: salt

    International Nuclear Information System (INIS)

    1978-04-01

    This volume, Volume 8 ''Repository Preconceptual Design Studies: Salt,'' is one of a 23-volume series, ''Technical Support for GEIS: Radioactive Waste Isolation in Geologic Formations,'' Y/OWI/TM-36, which supplements the ''Contribution to Draft Generic Environmental Impact Statement on Commercial Waste Management: Radioactive Waste Isolation in Geologic Formations,'' Y/OWI/TM-44. The series provides a more complete technical basis for the preconceptual designs, resource requirements, and environmental source terms associated with isolating commercial LWR wastes in underground repositories in salt, granite, shale and basalt. Wastes are considered from three fuel cycles: uranium and plutonium recycling, no recycling of spent fuel and uranium-only recycling. This document describes a preconceptual design for a nuclear waste storage facility in salt. The facility design consists of several chambers excavated deep within a geologic formation together with access shafts and supportive surface structures. The facility design provides for: receiving and unloading waste containers; lowering them down shafts to the mine level; transporting them to the proper storage area, and emplacing them in mined storage rooms. Drawings of the facility design are contained in TM-36/9, ''Drawings for Repository Preconceptual Design Studies: Salt.''

  20. Review of geochemical measurement techniques for a nuclear waste repository in bedded salt

    International Nuclear Information System (INIS)

    Knauss, K.G.; Steinborn, T.L.

    1980-01-01

    A broad, general review is presented of geochemical measurement techniques that can provide data necessary for site selection and repository effectiveness assessment for a radioactive waste repository in bedded salt. The available measurement techniques are organized according to the parameter measured. The list of geochemical parameters include all those measurable geochemical properties of a sample whole values determine the geochemical characteristics or behavior of the system. For each technique, remarks are made pertaining to the operating principles of the measurement instrument and the purpose for which the technique is used. Attention is drawn to areas where further research and development are needed

  1. The source term and waste optimization of molten salt reactors with processing

    International Nuclear Information System (INIS)

    Gat, U.; Dodds, H.L.

    1993-01-01

    The source term of a molten salt reactor (MSR) with fuel processing is reduced by the ratio of processing time to refueling time as compared to solid fuel reactors. The reduction, which can be one to two orders of magnitude, is due to removal of the long-lived fission products. The waste from MSRs can be optimized with respect to its chemical composition, concentration, mixture, shape, and size. The actinides and long-lived isotopes can be separated out and returned to the reactor for transmutation. These features make MSRs more acceptable and simpler in operation and handling

  2. High level nuclear waste repository in salt: Sealing systems status and planning report: Draft report

    International Nuclear Information System (INIS)

    1985-09-01

    This report documents the initial conceptual design studies for a repository sealing system for a high-level nuclear waste repository in salt. The first step in the initial design studies was to review the current design level, termed schematic designs. This review identified practicality of construction and development of a design methodology as two key issues for the conceptual design. These two issues were then investigated during the initial design studies for seal system materials, seal placement, backfill emplacement, and a testing and monitoring plan. The results of these studies have been used to develop a program plan for completion of the sealing system conceptual design. 60 refs., 26 figs., 18 tabs

  3. Aspects on the gas generation and migration in repositories for high level waste in salt formations

    International Nuclear Information System (INIS)

    Ruebel, Andre; Buhmann, Dieter; Meleshyn, Artur; Moenig, Joerg; Spiessl, Sabine

    2013-07-01

    In a deep geological repository for high-level waste, gases may be produced during the post-closure phase by several processes. The generated gases can potentially affect safety relevant features and processes of the repository, like the barrier integrity, the transport of liquids and gases in the repository and the release of gaseous radionuclides from the repository into the biosphere. German long-term safety assessments for repositories for high-level waste in salt which were performed prior 2010 did not explicitly consider gas transport and the consequences from release of volatile radionuclides. Selected aspects of the generation and migration of gases in repositories for high-level waste in a salt formation are studied in this report from the viewpoint of the performance assessment. The knowledge on the availability of water in the repository, in particular the migration of salt rock internal fluids in the temperature field of the radioactive waste repository towards the emplacement drifts, was compiled and the amount of water was roughly estimated. Two other processes studied in this report are on the one hand the release of gaseous radionuclides from the repository and their potential impact in the biosphere and on the other hand the transport of gases along the drifts and shafts of the repository and their interaction with the fluid flow. The results presented show that there is some gas production expected to occur in the repository due to corrosion of container material from water disposed of with the backfill and inflowing from the host rock during the thermal phase. If not dedicated gas storage areas are foreseen in the repository concept, these gases might exceed the storage capacity for gases in the repository. Consequently, an outflow of gases from the repository could occur. If there are failed containers for spent fuel, radioactive gases might be released from the containers into the gas space of the backfill and subsequently transported together

  4. Independent Assessment of the Savannah River Site High-Level Waste Salt Disposition Alternatives Evaluation

    International Nuclear Information System (INIS)

    Case, J. T.; Renfro, M. L.

    1998-01-01

    This report presents the results of the Independent Project Evaluation (IPE) Team assessment of the Westinghouse Savannah River Company High-Level Waste Salt Disposition Systems Engineering (SE) Team's deliberations, evaluations, and selections. The Westinghouse Savannah River Company concluded in early 1998 that production goals and safety requirements for processing SRS HLW salt to remove Cs-137 could not be met in the existing In-Tank Precipitation Facility as currently configured for precipitation of cesium tetraphenylborate. The SE Team was chartered to evaluate and recommend an alternative(s) for processing the existing HLW salt to remove Cs-137. To replace the In-Tank Precipitation process, the Savannah River Site HLW Salt Disposition SE Team down-selected (October 1998) 140 candidate separation technologies to two alternatives: Small-Tank Tetraphenylborate (TPB) Precipitation (primary alternative) and Crystalline Silicotitanate (CST) Nonelutable Ion Exchange (backup alternative). The IPE Team, commissioned by the Department of Energy, concurs that both alternatives are technically feasible and should meet all salt disposition requirements. But the IPE Team judges that the SE Team's qualitative criteria and judgments used in their down-selection to a primary and a backup alternative do not clearly discriminate between the two alternatives. To properly choose between Small-Tank TPB and CST Ion Exchange for the primary alternative, the IPE Team suggests the following path forward: Complete all essential R and D activities for both alternatives and formulate an appropriate set of quantitative decision criteria that will be rigorously applied at the end of the R and D activities. Concurrent conceptual design activities should be limited to common elements of the alternatives

  5. Possible salt mine and brined cavity sites for radioactive waste disposal in the northeastern southern peninsula of Michigan

    International Nuclear Information System (INIS)

    Landes, K.K.; Bourne, H.L.

    1976-01-01

    A reconnaissance report on the possibilities for disposal of radioactive waste covers Michigan only, and is more detailed than an earlier one involving the northeastern states. Revised ''ground rules'' for pinpointing both mine and dissolved salt cavern sites for waste disposal include environmental, geologic, and economic factors. The Michigan basin is a structural bowl of Paleozoic sediments resting on downwarped Precambrian rocks. The center of the bowl is in Clare and Gladwin Counties, a short distance north of the middle of the Southern Peninsula. The strata dip toward this central area, and some stratigraphic sequences, including especially the salt-containing Silurian section, increase considerably in thickness in that direction. Lesser amounts of salt are also present in the north central part of the Lower Peninsula. Michigan has been an oil and gas producing state since 1925 and widespread exploration has had two effects on the selection of waste disposal sites: (1) large areas are leased for oil and gas; and (2) the borehole concentrations, whether producing wells, dry holes, or industrial brine wells that penetrated the salt section, should be avoided. Two types of nuclear waste, low level and high level, can be stored in man-made openings in salt beds. The storage facilities are created by (1) the development of salt mines where the depths are less than 3000 ft, and (2) cavities produced by pumping water into a salt bed, and bringing brine back out. The high level waste disposal must be confined to mines of limited depth, but the low level wastes can be accommodated in brine cavities at any depth. Seven potential prospects have been investigated and are described in detail

  6. Preliminary area selection considerations for radioactive waste repositories in bedded salt

    International Nuclear Information System (INIS)

    Wagoner, J.L.; Steinborn, T.L.

    1979-01-01

    This guide describes an approach to selection of areas of bedded salt which contain potentially suitable sites for the storage of radioactive waste. To evaluate a site selected by a license applicant, it is necessary to understand the technical site characteristics which should be considered in the preliminary phase of site selection. These site characteristics are presented here in checklist form, and each item is accompanied by a discussion which explains its significance. These qualitative considerations are used first to select an area of interest within a broad geologic or geomorphic region. Once an area has been selected, more quantitative information must be acquired to determine whether the proposed site meets the resultations for storage of nuclear waste

  7. Modeling the dissolution behavior of defense waste glass in a salt repository environment

    International Nuclear Information System (INIS)

    McGrain, B.P.

    1988-02-01

    A mechanistic model describing a dynamic mass balance between the production and consumption of dissolved silica was found to describe the dissolution behavior of SRL-165 defense waste glass in a high-magnesium brine (PBB3) at a temperature of 90 0 C. The synergistic effect of the waste package container on the glass dissolution rate was found to depend on a precipitation reaction for a ferrous silicate mineral. The model predicted that the ferrous silicate precipitate should be variable in composition where the iron/silica stoichiometry depended on the metal/glass surface area ratio used in the experiment. This prediction was confirmed experimentally by the variable iron/silica ratios observed in filtered leachates. However, the interaction between dissolved silica and iron corrosion products needs to be much better understood before the model can be used with confidence in predicting radionuclide release rates for a salt repository. 25 refs., 4 figs., 1 tab

  8. Supplemental technical information in support of Y/OWI/TM--44. Volume 17. Drawings for repository preconceptual design studies: BPNL waste forms in salt

    International Nuclear Information System (INIS)

    1978-04-01

    Volume 17 contains drawings of a preconceptual design for a nuclear waste storage facility in salt. Three full cycles are considered: full recycle, throwaway cycle, and uranium recycle with plutonium in high-level waste

  9. Characterization and Potential Use of Biochar for the Remediation of Coal Mine Waste Containing Efflorescent Salts

    Directory of Open Access Journals (Sweden)

    Luis Carlos Díaz Muegue

    2017-11-01

    Full Text Available In open pit coal mining, soil and vegetation are removed prior to the start of mining activities, causing physical, chemical, and microbiological changes to the soil and landscape. The present work shows the results of an integrated study of the remediation of mine waste with a high level of salt contamination in areas of the Cesar Department (Colombia, employing biochar as an amendment. Physical-chemical properties including Munsell color, texture, pH, electrical conductivity, water-holding capacity, cation exchange capacity, metal content, organic carbon, sulfates, extractable P, and total nitrogen were characterized both in the soils contaminated with mine residues and the biochar sample. A high concentration of sulfates, calcium, iron, and aluminum and a significant presence of Na, followed by minor amounts of Mg, K, Cu, and Mn, were observed in efflorescent salts. X-ray diffraction indicated a high presence of quartz and gypsum and the absence of pyrite and Schwertmannite in the efflorescent salt, while showing broad peaks belonging to graphene sheets in the biochar sample. Soil remediation was evaluated in Petri dish seed germination bioassays using Brachiaria decumbens. Biochar was shown to be effective in the improvement of pH, and positively influenced the germination percentage and root length of Brachiaria grass seeds.

  10. Principal aspects of petrographical examination of rock salts to assess their suitability for radioactive waste disposal

    International Nuclear Information System (INIS)

    Shekhunova, S.B.

    1995-01-01

    To solve the problem of high-level radioactive waste (HLRW) isolation in Ukraine a preparatory stage of feasibility study as to the construction of a pilot laboratory has been completed. Salty formations are considered as possible host rocks for HLRW isolation. 7 salt formations located in 5 regions of Ukraine have been examined and was found that only two, i.e. the Upper Devonian and Lower Permian halogenic formations of the Dnieper-Donets Depression appeared to have considerable promise for these purposes. In these two formations 4 zones with 12 candidate-sites were selected. The promising zones are located both in bedded salt and in salt domes. Analytical treatment our previous studies as well as a special-purpose research have resulted in designing packages of the schematic information models for the zones and some candidate-sites. Now we are preparing to start exploration drilling at several promising structures. Research has been carried out by the Institute of Geological Sciences (National Academy of Sciences of Ukraine) on budget and contract financial basis with the participation of branch institutes and the State Committee on Nuclear power Utilization (Goskomatom). The drilling and geophysical data were presented by Goskomgeologiya production organizations

  11. Corrosion of carbon steel in saturated high-level waste salt solutions

    International Nuclear Information System (INIS)

    Wiersma, B.J.; Parish, W.R.

    1997-01-01

    High level waste stored as crystallized salts is to be removed from carbon steel tanks by water dissolution. Dissolution of the saltcake must be performed in a manner which will not impact the integrity of the tank. Corrosion testing was performed to determine the amount of corrosion inhibitor that must be added to the dissolution water in order to ensure that the salt solution formed would not induce corrosion degradation of the tank materials. The corrosion testing performed included controlled potential slow strain rate, coupon immersion, and potentiodynamic polarization tests. These tests were utilized to investigate the susceptibility of the cooling coil material to stress corrosion cracking in the anticipated environments. No evidence of SCC was observed in any of the tests. Based on these results, the recommended corrosion requirements were that the temperature of the salt solution be less than 50 degrees C and that the minimum hydroxide concentration be 0.4 molar. It was also recommended that the hydroxide concentration not stay below 0.4 molar for longer than 45 days

  12. Final status of the salt repository project waste package program experimental database

    International Nuclear Information System (INIS)

    Thornton, B.M.; Reimus, P.W.

    1988-03-01

    This report describes the final status of the Salt Repository Project Waste Package Program Experimental Database. The data base serves as a clearinghouse for all data collected within the Waste Package Program (WPP) and its predecessor programs at Pacific Northwest Laboratory (PNL). The database was maintained using RS/1 database management software. Documented assurance that the entries in the database were consistent with experimental records was provided by having each experimentalist inspect the entries and signify that they were in agreement with the records. The inspection and signoff were done per PNL technical procedures. Data for which it was impossible to obtain the experimentalist's inspection and signature were segregated from the rest of the database, although they could still be accessed by WPP staff. The WPPED contains two groups of subdirectories. One group contains data taken prior to the installation of quality assurance procedures at PNL. The other group of subdirectories contains data taken under the NQA-1 procedures since their installation in April 1985. As part of closeout activities in the Salt Repository Project, the WPP database has been archived onto magnetic media. The data in the database are available by request on magnetic media or in hardcopy form. 2 refs

  13. Trial storage of high-level waste cylinders in the Asse II salt mine

    International Nuclear Information System (INIS)

    1984-01-01

    This report covers the contract period 1976-77, as well as some of the tasks carried out during the extension in 1978, in the framework of the R and D programme for disposal of radioactive waste in salt formations. With regard to the in-situ tests for the liberation and migration of brine, the testing devices were examined successfully. Laboratory examinations carried out showed a stepwise liberation of the water contents in halite in dependence on the temperature. The amount of brine liberated stood in good agreement with the in situ results. A temperature test for borehole convergence resulted in definite convergence rates. Simultaneously no influence was registered in the stability of the surrounding rocks. For the realization of an integrated major experiment, temperature test field IV was mined on the 750 m level of the Asse Salt Mine and heater- as well as measurement drillings were carried out. Extensive rheological examinations are concentrated particularly on the halite and secondly on the Carnallite. They are chiefly based on uni- and multiaxial pressure tests. Computer programmes are developed to examine the heat generation in wastes as well as in salt. In comparison, the programme development of computer codes for the stability behaviour of rocks is still at a relatively early stage, because it has to build up on the results of heat generation. The works for the development of a transport container with a shielding combination are at a very advanced stage. An integrated disposal- and retrieval system was developed, tested and successfully demonstrated. A monitoring system in the mine has also been developed in its essential parts

  14. Radioactive waste isolation in salt: peer review of Office of Nuclear Waste Isolation's Socioeconomic Program Plan

    International Nuclear Information System (INIS)

    Winter, R.; Fenster, D.; O'Hare, M.; Zillman, D.; Harrison, W.; Tisue, M.

    1984-07-01

    The following recommendations have been abstracted from the body of this report. The Office of Nuclear Waste Isolation's Socioeconomic Program Plan for the Establishment of Mined Geologic Repositories to Isolate Nuclear Waste should be modified to: (1) encourage active public participation in the decision-making processes leading to repository site selection; (2) clearly define mechanisms for incorporating the concerns of local residents, state and local governments, and other potentially interested parties into the early stages of the site selection process. In addition, the Office of Nuclear Waste Isolation should carefully review the overall role that these persons and groups, including local pressure groups organized in the face of potential repository development, will play in the siting process; (3) place significantly greater emphasis on using primary socioeconomic data during the site selection process, reversing the current overemphasis on secondary data collection, description of socioeconomic conditions at potential locations, and development of analytical methodologies; (4) include additional approaches to solving socioeconomic problems. For example, a reluctance to acknowledge that solutions to socioeconomic problems need to be found jointly with interested parties is evident in the plan; (5) recognize that mitigation mechanisms other than compensation and incentives may be effective; (6) as soon as potential sites are identified, the US Department of Energy (DOE) should begin discussing impact mitigation agreements with local officials and other interested parties; and (7) comply fully with the pertinent provisions of NWPA

  15. Removal of salt from high-level waste tanks by density-driven circulation or mechanical agitation

    International Nuclear Information System (INIS)

    Kiser, D.L.

    1981-01-01

    Twenty-two high-level waste storage tanks at the Savannah River Plant are to be retired in the tank replacement/waste transfer program. The salt-removal portion of this program requires dissolution of about 19 million liters of salt cake. Steam circulation jets were originally proposed to dissolve the salt cake. However, the jets heated the waste tank to 80 to 90 0 C. This high temperature required a long cooldown period before transfer of the supernate by jet, and increased the risk of stress-corrosion cracking in these older tanks. A bench-scale investigation at the Savannah River Laboratory developed two alternatives to steam-jet circulation. One technique was density-driven circulation, which in bench tests dissolved salt at the same rate as a simulated steam circulation jet but at a lower temperature. The other technique was mechanical agitation, which dissolved the salt cake faster and required less fresh water than either density-driven circulation or the simulated steam circulation jet. Tests in an actual waste tank verified bench-scale results and demonstrated the superiority of mechanical agitation

  16. Recent studies on radiation damage formation in synthetic NaCl and natural rock salt for radioactive waste disposal applications

    International Nuclear Information System (INIS)

    Swyler, K.J.; Klaffky, R.W.; Levy, P.W.

    1980-01-01

    Radiation damage formation in natural rock salt is described as a function of irradiation temperature and plastic deformation. F-center formation decreases with increasing temperature while significant colloidal sodium formation occurs over a restricted temperature range around 150 0 C. Plastic deformation increases colloid formation; it is estimated that colloid concentrations may be increased by a factor of 3 if the rock salt near radioactive waste disposal canisters is heavily deformed. Optical bandshape analysis indicates systematic differences between the colloids formed in synthetic and natural rock salts

  17. Radioactive waste isolation in salt: Peer review of the Golder Associates draft test plan for in situ testing in an exploratory shaft in salt

    International Nuclear Information System (INIS)

    Hambley, D.F.; Mraz, D.Z.; Unterberter, R.R.

    1987-01-01

    This report documents the peer review conducted by Argonne National Laboratory of a document entitled ''Draft Test Plan for In Situ Testing in an Exploratory Shaft in Salt,'' prepared for Battelle Memorial Institute's Office of Nuclear Waste Isolation by Golder Associates, Inc. In general, the peer review panelists found the test plan to be technically sound, although some deficiencies were identified. Recommendations for improving the test plan are presented in this review report. A microfiche copy of the following unpublished report is attached to the inside back cover of this report: ''Draft Test Plan for In Situ Testing in an Exploratory Shaft in Salt,'' prepared by Golder Associates, Inc., for Office of Nuclear Waste Isolation, Battelle Memorial Institute, Columbus, Ohio (March 1985)

  18. Technetium removal column flow testing with alkaline, high salt, radioactive tank waste

    International Nuclear Information System (INIS)

    Blanchard, D.L. Jr.; Kurath, D.E.; Golcar, G.R.; Conradson, S.D.

    1996-01-01

    This report describes two bench-scale column tests conducted to demonstrate the removal of Tc-99 from actual alkaline high salt radioactive waste. The waste used as feed for these tests was obtained from the Hanford double shell tank AW-101, which contains double shell slurry feed (DSSF). The tank sample was diluted to approximately 5 M Na with water, and most of the Cs-137 was removed using crystalline silicotitanates. The tests were conducted with two small columns connected in series, containing, 10 mL of either a sorbent, ABEC 5000 (Eichrom Industries, Inc.), or an anion exchanger Reillex trademark-HPQ (Reilly Industries, Inc.). Both materials are selective for pertechnetate anion (TcO 4 - ). The process steps generally followed those expected in a full-scale process and included (1) resin conditioning, (2) loading, (3) caustic wash to remove residual feed and prevent the precipitation of Al(OH) 3 , and (4) elution. A small amount of Tc-99m tracer was added as ammonium pertechnetate to the feed and a portable GEA counter was used to closely monitor the process. Analyses of the Tc-99 in the waste was performed using ICP-MS with spot checks using radiochemical analysis. Technetium x-ray absorption spectroscopy (XAS) spectra of 6 samples were also collected to determine the prevalence of non-pertechnetate species [e.g. Tc(IV)

  19. A general model for the dissolution of nuclear waste glasses in salt brine

    International Nuclear Information System (INIS)

    McGrail, B.P.; Strachan, D.M.

    1988-07-01

    A mechanistic model describing a dynamic mass balance between the production and consumption of dissolved silica was found to describe the dissolution of SRL-165 defense waste glass in a high-magnesium (PBB3) brine at a temperature of 90/degree/C. The synergetic effect of the waste package container on the glass dissolution rate was found to depend on a precipitation reaction for a ferrous silicate mineral. The model predicted that the ferrous silicate precipitate should be variable in composition where the iron-silica ratio depended on the metal-to-glass surface area ratio used in the experiment. This prediction was confirmed experimentally by the variable iron-silica ratios observed in filtered leachates. However, the interaction between dissolved silica and iron corrosion products needs to be much better understood before the model could be used with confidence in predicting radionuclide release rates for a salt repository. If the deleterious effects of the iron corrosion products can be shown to be transient, and the fracturing of the glass can be minimized, it appears that the performance of SRL-165 defense waste glass will be near the NRC regulatory criterion for fraction release of one part in 100,000 in PBB3 brine at 90/degree/C under silica-saturated conditions. 47 refs., 6 figs., 1 tab

  20. On the time-dependent behavior of a cylindrical salt dome with a high-level waste repository

    International Nuclear Information System (INIS)

    Prij, J.

    1988-01-01

    In a salt dome with a repository for high-level radioactive and heat-generating waste, thermal stresses develop. These stresses can influence the isolation capability of the salt dome if these stresses can initiate cracks or introduce movements along existing closed flaws. The influence of the thermomechanical properties of the rock salt and the surrounding rocks on the thermal stresses and the surface rise is discussed. This discussion is based on a number of finite element creep analyses of a homogeneous cylindrical salt dome. The parameters, varied in the analyses, are constants in the thermomechanical constitutive behavior of salt and rocks, and furthermore the thermal loading has been varied. It is shown that variations in the creep properties, which result in differences in creep strain rate of a factor of 100, have only a very limited influence on the thermal stresses and the surface rise. Of more importance is the elastic stiffness of the materials. In all creep analyses the thermal stresses in the salt are compressive and the shear stresses remain below 2 MPa. The results are evaluated using an analytical treatment. Based on this evaluation, it is shown that the observed trends in the numerical results have a more general character and are not strictly limited to the geometry chosen. It is concluded that the thermal stresses in the salt formation are not strongly dependent on the creep properties of the rock salt

  1. Radioactive waste isolation in salt: peer review of Office of Nuclear Waste Isolation's Socioeconomic Program Plan

    International Nuclear Information System (INIS)

    Winter, R.; Fenster, D.; O'Hare, M.; Zillman, D.; Harrison, W.; Tisue, M.

    1984-02-01

    The ONWI Socioeconomic Program Plan spells out DOE's approach to analyzing the socioeconomic impacts from siting, constructing, and operating radioactive waste repositories and discusses mitigation strategies. The peer review indicated the following modifications should be made to the Plan: encourage active public participation in the decision-making processes leading to repository site selection; clearly define mechanisms for incorporating the concerns of local residents, state and local governments, and other potentially interested parties into the early stages of the site selection process; place significantly greater emphasis on using primary socioeconomic data during the site selection process, reversing the current overemphasis on secondary data collection, description of socioeconomic conditions at potential locations, and development of analytical methodologies; recognize that mitigation mechanisms other than compensation and incentives may be effective; as soon as potential sites are identified, the US Department of Energy (DOE) should begin discussing impact mitigation agreements with local officials and other interested parties; and comply fully with the pertinent provisions of NWPA

  2. Molten salt oxidation of mixed wastes: Separation of radioactive materials and Resource Conservation and Recovery Act (RCRA) materials

    International Nuclear Information System (INIS)

    Bell, J.T.; Haas, P.A.; Rudolph, J.C.

    1993-01-01

    The Oak Ridge National Laboratory (ORNL) is involved in a program to apply a molten salt oxidation (MSO) process to the treatment of mixed wastes at Oak Ridge and other Department of Energy (DOE) sites. Mixed wastes are defined as those wastes that contain both radioactive components, which are regulated by the atomic energy legislation, and hazardous waste components, which are regulated under the Resource Conservation and Recovery Act (RCRA). A major part of our ORNL program involves the development of separation technologies that are necessary for the complete treatment of mixed wastes. The residues from the MSO treatment of the mixed wastes must be processed further to separate the radioactive components, to concentrate and recycle residues, or to convert the residues into forms acceptable for final disposal. This paper is a review of the MSO requirements for separation technologies, the information now available, and the concepts for our development studies

  3. Geochemical processes in marine salt deposits: Their significance and their implications in connection with disposal of radioactive waste within salt domes

    Energy Technology Data Exchange (ETDEWEB)

    Herrmann, A G [Goettingen Univ. (Germany, F.R.). Geochemisches Inst.

    1980-01-01

    Attempts to effect permanent disposal of radioactive wastes in marine evaporites should do nothing to disturb, either in the short or the long term, the present relative stability of such bodies of rock. It is necessary to take account of all of the geochemical and physico-chemical reactions known to have been involved in the processes which formed the evaporites before proceeding to an acceptable strategy for disposal of radionucleides. These processes can be represented as three kinds of metamorphism: 1. solution metamorphism, 2. thermal metamorphism, 3. dynamic metamorphism. In all of the evaporite occurrences in Germany such processes have been influential in altering, on occasion significantly, the primary mineralogical composition and have also promoted a considerable degree of transposition of material. Given similar geochemical and physico-chemical premises, these metamorphic processes could become effective now or in the future. It is therefore necessary to discuss the following criteria when examining salt domes as permanent repositories of highly radioactive substances: (1) Temperatures <= 90/sup 0/ +- 10/sup 0/C at the contact between waste containers and rock salt; (2) Temperatures <= 75/sup 0/C within zones of carnallite rocks; (3) Immobilisation of high-level waste in crystalline forms whenever possible; (4) Systems of additional safety barriers around the waste containers or the unreprocessed spent fuel elements. The geochemical and physical effectiveness of the barriers within an evaporite environment must be guaranteed. For example: Ni-Ti-alloys, corundum, ceramic, anhydrite.

  4. Hydrothermal preparation of zeolite Li-A and ion exchange properties of Cs and Sr in salt waste

    International Nuclear Information System (INIS)

    Lee, S. H.; Kim, J. G.; Lee, J. H.; Kim, J. H.

    2005-01-01

    An advanced spent fuel management process that were based on Li reduction of the oxide spent fuel to a metallic form will generate a LiCl waste. Zeolite A has been reported as a promising immobilization medium for waste salt with CsCl and SrCl 2 . However, Sodium is accumulated as an ionic form (Na + -ion) in molten salt during ion exchange step between Na + -ion in zeolite A and Li + -ion in the molten salt. Therefore, zeolite Na-A need to be replaced by the Li-type zeolite for recycling the salt waste by removing the Cs and Sr ions. In this study, the hydrothermal preparation of zeolite Li-A was performed in 350ml pressure vessel by P. Norby method. The preparation characteristics of zeolite Li-A was investigated. And the ion exchange properties of Cs and Sr in molten LiCl salt were investigated under the condition of 923K using zeolite 4A and prepared zeolite Li-A

  5. Permeability of natural rock salt from the Waste Isolation Pilot Plant (WIPP) during damage evolution and healing

    International Nuclear Information System (INIS)

    Pfeifle, T.W.; Hurtado, L.D.

    1998-06-01

    The US Department of Energy has developed the Waste Isolation Pilot Plant (WIPP) in the bedded salt of southeastern New Mexico to demonstrate the safe disposal of radioactive transuranic wastes. Four vertical shafts provide access to the underground workings located at a depth of about 660 meters. These shafts connect the underground facility to the surface and potentially provide communication between lithologic units, so they will be sealed to limit both the release of hazardous waste from and fluid flow into the repository. The seal design must consider the potential for fluid flow through a disturbed rock zone (DRZ) that develops in the salt near the shafts. The DRZ, which forms initially during excavation and then evolves with time, is expected to have higher permeability than the native salt. The closure of the shaft openings (i.e., through salt creep) will compress the seals, thereby inducing a compressive back-stress on the DRZ. This back-stress is expected to arrest the evolution of the DRZ, and with time will promote healing of damage. This paper presents laboratory data from tertiary creep and hydrostatic compression tests designed to characterize damage evolution and healing in WIPP salt. Healing is quantified in terms of permanent reduction in permeability, and the data are used to estimate healing times based on considerations of first-order kinetics

  6. Cerebral salt-wasting syndrome after hematopoietic stem cell transplantation in adolescents: 3 case reports

    Directory of Open Access Journals (Sweden)

    Yeon Jin Jeon

    2015-12-01

    Full Text Available Cerebral salt-wasting syndrome (CSWS is a rare disease characterized by a extracellular volume depletion and hyponatremia induced by marked natriuresis. It is mainly reported in patients who experience a central nervous system insult, such as cerebral hemorrhage or encephalitis. The syndrome of inappropriate antidiuretic hormone secretion is a main cause of severe hyponatremia after hematopoietic stem cell transplantation, whereas CSWS is rarely reported. We report 3 patients with childhood acute leukemia who developed CSWS with central nervous system complication after hematopoietic stem cell transplantation. The diagnosis of CSW was made on the basis of severe hyponatremia accompanied by increased urine output with clinical signs of dehydration. All patients showed elevated natriuretic peptide and normal antidiuretic hormone. Aggressive water and sodium replacement treatment was instituted in all 3 patients and 2 of them were effectively recovered, the other one was required to add fludrocortisone administration.

  7. Review of geotechnical measurement techniques for a nuclear waste repository in bedded salt

    International Nuclear Information System (INIS)

    1979-12-01

    This report presents a description of geotechnical measurement techniques that can provide the data necessary for safe development - i.e., location, design, construction, operation, decommissioning and abandonment - of a radioactive waste repository in bedded salt. Geotechnical data obtained by a diversity of measurement techniques are required during all phases of respository evolution. The techniques discussed in this report are grouped in the following categories: geologic, geophysical and geodetic; rock mechanics; hydrologic, hydrogeologic and water quality; and thermal. The major contribution of the report is the presentation of extensive tables that provide a review of available measurement techniques for each of these categories. The techniques are also discussed in the text to the extent necessary to describe the measurements and associated instruments, and to evaluate the applicability or limitations of the method. More detailed discussions of thermal phenomena, creep laws and geophysical methods are contained in the appendices; references to detailed explanations of measurement techniques and instrumentation are inluded throughout the report

  8. Review of geotechnical measurement techniques for a nuclear waste repository in bedded salt

    Energy Technology Data Exchange (ETDEWEB)

    1979-12-01

    This report presents a description of geotechnical measurement techniques that can provide the data necessary for safe development - i.e., location, design, construction, operation, decommissioning and abandonment - of a radioactive waste repository in bedded salt. Geotechnical data obtained by a diversity of measurement techniques are required during all phases of respository evolution. The techniques discussed in this report are grouped in the following categories: geologic, geophysical and geodetic; rock mechanics; hydrologic, hydrogeologic and water quality; and thermal. The major contribution of the report is the presentation of extensive tables that provide a review of available measurement techniques for each of these categories. The techniques are also discussed in the text to the extent necessary to describe the measurements and associated instruments, and to evaluate the applicability or limitations of the method. More detailed discussions of thermal phenomena, creep laws and geophysical methods are contained in the appendices; references to detailed explanations of measurement techniques and instrumentation are inluded throughout the report.

  9. Towards controlling dioxins emissions from power boilers fuelled with salt-laden wood waste

    International Nuclear Information System (INIS)

    Luthe, C.; Karidio, I.; Uloth, V.

    1997-01-01

    An evaluation of the dioxins emissions from a power boiler fuelled with salt-laden wood waste has provided insights on potential control technologies. Whereas a reduction in stack particulate levels does not preclude a corresponding reduction in dioxins emissions, good combustion conditions, in combination with an efficient secondary collection device for particulate removal, were found to offer effective control (stack emissions of 0.064 to 0.086 ng TEQ/m 3 ). Regarding minimization of dioxins formation at source, a preliminary assessment of the possible beneficial effect of an attenuated chlorine:sulphur ratio was encouraging. A more accurate assessment requires additional trials, preferably longer in duration, to eliminate any possible memory effects. (author)

  10. Groundwater recharge and discharge scenarios for a nuclear waste repository in bedded salt

    International Nuclear Information System (INIS)

    Carpenter, D.W.; Steinborn, T.L.; Thorson, L.D.

    1979-01-01

    Twelve potential scenarios have been identified whereby groundwater may enter or exit a nuclear waste repository in bedded salt. The 12 scenarios may be grouped into 4 categories or failure modes: dissolution, fracturing, voids, and penetration. Dissolution modes include breccia pipe and breccia blanket formation, and dissolution around boreholes. Fracture modes include flow through preexisting or new fractures and the effects of facies changes. Voids include interstitial voids (pores) and fluid inclusions. Penetration modes include shaft and borehole sealing failures, undetected boreholes, and new mines or wells constructed after repository decommissioning. The potential importance of thermal effects on groundwater flow patterns and on the recharge-discharge process is discussed. The appropriate levels of modeling effort, and the interaction between the adequacy of the geohydrologic data base and the warranted degree of model complexity are also discussed

  11. Performance assessment of geological isolation systems for radioactive waste. Disposal in salt formations

    International Nuclear Information System (INIS)

    Storck, R.; Aschenbach, J.; Hirsekom, R.P.; Nies, A.; Stelte, N.

    1988-01-01

    In the framework of the PAGIS project of the CEC Research Programme on radioactive waste, a performance assessment of a repository of vitrified HLW in rock salt formations has been carried out. The first volume of the study is split into four tasks. Task 1 recalls the main steps that have led to the selection of the reference and the variant site. Task 2 condenses all information available on the rock formations which are planned to host the repository, the overlying geosphere and the geohistoric development of the sites. Task 3 states the technical details of repository planning, while in Task 4 conceivable release scenarios are discussed. Volume II (Tasks 5 to 10) is concerned with the modelling procedures. In Task 5 data for the waste inventory are collected and the selection of relevant nuclides for transport calculations is discussed. Task 6 gives the near-field modelling, i.e. the models for corrosion of the waste canisters, the degradation of the waste matrix and the models used for the HLW boreholes. Task 7 deals with the modelling of the repository. Its division into sections is discussed and models for physical and chemical effects taken into account in each section are presented. In Task 8 the modelling of the overburden is given. In Task 9 additional models for the subrosion scenario and a human intrusion scenario are given. Task 10 is concerned with the biosphere modelling. In Volume III results of deterministic and probabilistic calculations are presented. Task 11 gives the results for deterministic calculations with best estimate values for the parameters involved in the models. Task 12 presents the result of the uncertainty analysis, and Task 13 those of local and global sensitivity analyses followed by concluding remarks. This document is one of a set of 5 reports covering a relevant project of the European Community on a nuclear safety subject having very wide interest. The five volumes are: the summary (EUR 11775-EN), the clay (EUR 11776-EN), the

  12. Delayed diagnosis of congenital adrenal hyperplasia with salt wasting due to type II 3beta-hydroxysteroid dehydrogenase deficiency

    DEFF Research Database (Denmark)

    Johannsen, Trine H; Mallet, Delphine; Dige-Petersen, Harriet

    2005-01-01

    Classical 3beta-hydroxysteroid dehydrogenase (3beta-HSD) deficiency is a rare cause of congenital adrenal hyperplasia. We report two sisters presenting with delayed diagnoses of classical 3beta-HSD, despite salt wasting (SW) episodes in infancy. Sibling 1 was referred for premature pubarche, slig...

  13. Estimates of relative areas for the disposal in bedded salt of LWR wastes from alternative fuel cycles

    International Nuclear Information System (INIS)

    Lincoln, R.C.; Larson, D.W.; Sisson, C.E.

    1978-01-01

    The relative mine-level areas (land use requirements) which would be required for the disposal of light-water reactor (LWR) radioactive wastes in a hypothetical bedded-salt formation have been estimated. Five waste types from alternative fuel cycles have been considered. The relative thermal response of each of five different site conditions to each waste type has been determined. The fuel cycles considered are the once-through (no recycle), the uranium-only recycle, and the uranium and plutonium recycle. The waste types which were considered include (1) unreprocessed spent reactor fuel, (2) solidified waste derived from reprocessing uranium oxide fuel, (3) plutonium recovered from reprocessing spent reactor fuel and doped with 1.5% of the accompanying waste from reprocessing uranium oxide fuel, (4) waste derived from reprocessing mixed uranium/plutonium oxide fuel in the third recycle, and (5) unreprocessed spent fuel after three recycles of mixed uranium/plutonium oxide fuels. The relative waste-disposal areas were determined from a calculated value of maximum thermal energy (MTE) content of the geologic formations. Results are presented for each geologic site condition in terms of area ratios. Disposal area requirements for each waste type are expressed as ratios relative to the smallest area requirement (for waste type No. 2 above). For the reference geologic site condition, the estimated mine-level disposal area ratios are 4.9 for waste type No. 1, 4.3 for No. 3, 2.6 for No. 4, and 11 for No. 5

  14. Effect of kefir and low-dose aspirin on arterial blood pressure measurements and renal apoptosis in unhypertensive rats with 4 weeks salt diet.

    Science.gov (United States)

    Kanbak, Güngör; Uzuner, Kubilay; Kuşat Ol, Kevser; Oğlakçı, Ayşegül; Kartkaya, Kazım; Şentürk, Hakan

    2014-01-01

    Abstract We aim to study the effect of low-dose aspirin and kefir on arterial blood pressure measurements and renal apoptosis in unhypertensive rats with 4 weeks salt diet. Forty adult male Sprague-Dawley rats were divided into five groups: control, high-salt (HS) (8.0% NaCl), HS+aspirin (10 mg/kg), HS+kefir (10.0%w/v), HS+aspirin +kefir. We measured sistolic blood pressure (SBP), mean arterial pressure (MAP), diastolic pressure, pulse pressure in the rats. Cathepsin B, L, DNA fragmentation and caspase-3 activities were determined from rat kidney tissues and rats clearance of creatinine calculated. Although HS diet increased significantly SBP, MAP, diastolic pressure, pulse pressure parameters compared the control values. They were not as high as accepted hypertension levels. When compared to HS groups, kefir groups significantly decrease Cathepsin B and DNA fragmentation levels. Caspase levels were elevated slightly in other groups according to control group. While, we also found that creatinine clearance was higher in HS+kefir and HS+low-dose aspirin than HS group. Thus, using low-dose aspirin had been approximately decreased of renal function damage. Kefir decreased renal function damage playing as Angiotensin-converting enzyme inhibitor. But, low-dose aspirin together with kefir worsened rat renal function damage. Cathepsin B might play role both apoptosis and prorenin-processing enzyme. But not caspase pathway may be involved in the present HS diet induced apoptosis. In conclusion, kefir and low-dose aspirin used independently protect renal function and renal damage induced by HS diet in rats.

  15. Creep tests on clean and argillaceous salt from the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    Mellegard, K.D.; Pfeifle, T.W.

    1993-05-01

    Fifteen triaxial compression creep tests were performed on clean and argillaceous salt from the Waste Isolation Pilot Plant (WIPP). The temperatures in the tests were either 25 degrees C or 100 degrees C while the stress difference ranged from 3.5 MPa to 21.0 MPa. In all tests, the confining pressure was 15 MPa. Test duration ranged from 23 to 613 days with an average duration of 300 days. The results of the creep tests supplemented earlier testing and were used to estimate two parameters in the Modified Munson-Dawson constitutive law for the creep behavior of salt. The two parameters determined from each test were the steady-state strain rate and the transient strain limit. These estimates were combined with parameter estimates determined from previous testing to study the dependence of both transient and steady-state creep deformation on stress difference. The exponents on stress difference determined in this study were found to be consistent with revised estimates of the exponents reported by other investigators

  16. Planned investigations for packing materials for a waste package in a salt repository: [Final report

    International Nuclear Information System (INIS)

    Shade, J.W.; Bunnell, L.R.; Thornton, T.A.

    1987-10-01

    A considerable number of materials have been either proposed or investigated as packing materials for nuclear waste package systems. Almost always the expandable clays, such as the smectites contained in commercial bentonites, have received the most attention when their primary function is to retard groundwater flow. Other materials including zeolites, metals, and dessicants are considered as special-purpose additives. Materials that tend to hydrolyze and lead to porosity reduction, such as silicates, oxides, and sulfates, have also been suggested as packing materials. All these types of materials are also considered as components of tailored mixtures to achieve a broad range of packing material performance. Some of these materials are reviewed, along with proposed candidate materials, with respect to the properties required to function in a salt repository. The investigation of packing materials is composed of five studies which are discussed below. Initial candidates will consist of calcium hydroxide, a sodium silicate, and a cement-gypsum mixture in addition to the reference crushed salt. Consequently these tests will be necessary to determine properties of individual components and to optimize properties of mixtures. 13 refs., 7 figs., 1 tab

  17. Heat transfer analysis of the waste-container sleeve/salt configuration

    International Nuclear Information System (INIS)

    Callahan, G.D.; Ratigan, J.L.; Russell, J.E.; Fossum, A.F.

    1975-01-01

    Prior to this investigation, the heat transport considered was only that of straight conduction. The waste container, air gap, and sleeve arrangement was considered to be a single, consistent, time-dependent, heat-generating unit in intimate contact with the salt. The conduction model does not accurately model the heat transfer mechanisms available. Thus radiation and combined radiation and convection must also be considered in the determination of the temperature field. As would be expected, the canister temperatures are higher for the case of radiation across the airgap than those that result from conduction when the canister is in intimate contact with the salt. For the radiation case, the canister temperatures rise rapidly to a temperature of approximately 1,140 0 F and maintain an almost steady state condition for one year whereafter the temperatures slowly decrease. The far field temperatures, near the pillar centerline, are essentially equivalent for all cases. As time proceeds, the far field temperatures of the conduction models are about 15% different

  18. Molten salt destruction as an alternative to open burning of energetic material wastes

    International Nuclear Information System (INIS)

    Upadhye, R.S.; Watkins, B.E.; Pruneda, C.O.; Brummond, W.A.

    1994-01-01

    LLNL has built a small-scale (about 1 kg/hr throughput unit to test the destruction of energetic materials using the Molten Salt Destruction (MSD) process. We have modified the unit described in the earlier references to inject energetic waste material continuously into the unit. In addition to the HMX, other explosives we have destroyed include RDX, PETN, ammonium picrate, TNT, nitroguanadine, and TATB. We have also destroyed a liquid gun propellant comprising hydroxyl ammonium nitrate, triethanolammonium nitrate and water. In addition to these pure components, we have destroyed a number of commonly used formulations, such as LX-10 (HMX/Viton), LX-16 (PETN/FPC461, LX-17 (TATB/Kel F), and PBX-9404 (HMX)/CEF/Nitro cellulose). Our experiments have demonstrated that energetic materials can be safely and effectively treated by MSD.We have also investigated the issue of steam explosions in molten salt units, both experimentally and theoretically, and concluded that steam explosions can be avoided under proper design and operating conditions. We are currently building a larger unit (nominal capacity 5 kg/hr,) to investigate the relationship between residence time, temperature, feed concentration and throughputs, avoidance of back-burn, a;nd determination of the products of combustion under different operating conditions

  19. Radioactive Waste Isolation in Salt: Peer review of documents dealing with geophysical investigations

    International Nuclear Information System (INIS)

    McGinnis, L.D.; Bowen, R.H.

    1987-03-01

    The Salt Repository Project, a US Department of Energy program to develop a mined repository in salt for high-level radioactive waste, is governed by a complex and sometimes inconsistent array of laws, administrative regulations, guidelines, and position papers. In conducting multidisciplinary peer reviews of contractor documents in support of this project, Argonne National Laboratory has needed to inform its expert reviewers of these governmental mandates, with particular emphasis on the relationship between issues and the technical work undertaken. This report acquaints peer review panelists with the regulatory framework as it affects their reviews of site characterization plans and related documents, including surface-based and underground test plans. Panelists will be asked to consider repository performance objectives and issues as they judge the adequacy of proposed geophysical testing. All site-specific discussions relate to the Deaf Smith County site in Texas, which was approved for site characterization by the President in May 1986. Natural processes active at the Deaf Smith County site and the status of geophysical testing near the site are reviewed briefly. 25 refs., 4 figs., 5 tabs

  20. Disposal of high-level waste from nuclear power plants in Denmark. Salt dome investigations. v.5

    International Nuclear Information System (INIS)

    1981-01-01

    The present report deals with safety evaluation as part of the investigations regarding a repository for high-level waste in a salt dome. It is volume 5 of five volumes that together constitute the final report on the Danish utilities' salt dome investigations. Two characteristics of the waste are of special importance for the safety evaluation: the encasing of the waste in steel casks with 15 cm thick walls affording protection against corrosion, protecting the surroundings against radiation, and protecting the glass cylinders from mechanical damage resulting from the pressure at the bottom of the disposal hole, and the modest generation of heat in the waste at the time of disposal resulting in a maximum temperature increase in the salt close to the waste of approx. 40 deg. C. These characteristics proved to considerably improve the safety margin with respect to unforeseen circumstances. The character of the salt dome and of the salt in the proposed disposal area offers in itself good protection against contact with the ground water outside the dome. The relatively large depth of 1200 and 2500 m of the salt surface also means that neither dome nor disposal facility will be appreciably influenced by glaciations or earthquakes. The chalk above the proposed disposal area is very tight and to retain radioactive matter effectively even in the precence of high concentrations of NaCL. The safety investigations included a number of natural processes and probable events such as the segregation of crystal water from overlooked salt minerals, faulty sealings of disposal holes, permeable fault zones in the chalk overlying the dome, the risk in connection with human penetration into the dome. These conditions will neither lead to the destruction of the waste casks or to the release of waste from the dome. Leaching of a cavern is the only situation which proved to result in a release of radioactive material to the biosphere, but the resulting doses was found to be small

  1. Safety assessment of radioactive waste disposal into geological formations; a preliminary application of fault tree analysis to salt deposits

    International Nuclear Information System (INIS)

    Bertozzi, B.; D'Alessandro, M.; Girardi, F.; Vanossi, M.

    1978-01-01

    The methodology of the fault tree analysis (FTA) has been widely used at the Joint Research Centre of Ispra in nuclear reactor safety studies. The aim of the present work consisted in studying the applicability of this methodology to geological repositories of radioactive wastes, including criteria and approaches for the quantification of probalities of primary events. The present work has just an illustrative purpose. Two ideal cases of saline formations, I.E. a bedded salt and a diapir were chosen as potential disposal sites for radioactive waste. On the basis of arbitrarily assumed hydrogeological features of the salt formations and their surrounding environment, possible phenomena capable of causing the waste to be released from each formation have been discussed and gathered following the logical schemes of the FTA. The assessment of probability values for release events due to natural causes as well as to human actions, over different time periods, up to one million years, has been discussed

  2. Conceptual design of retrieval systems for emplaced transuranic waste containers in a salt bed depository. Final report

    International Nuclear Information System (INIS)

    Fogleman, S.F.

    1980-04-01

    The US Department of Energy and the Nuclear Regulatory Commission have jurisdiction over the nuclear waste management program. Design studies were previously made of proposed repository site configurations for the receiving, processing, and storage of nuclear wastes. However, these studies did not provide operational designs that were suitable for highly reliable TRU retrieval in the deep geologic salt environment for the required 60-year period. The purpose of this report is to develop a conceptual design of a baseline retrieval system for emplaced transuranic waste containers in a salt bed depository. The conceptual design is to serve as a working model for the analysis of the performance available from the current state-of-the-art equipment and systems. Suggested regulations would be based upon the results of the performance analyses

  3. Impact of Salt Waste Processing Facility Streams on the Nitric-Glycolic Flowsheet in the Chemical Processing Cell

    Energy Technology Data Exchange (ETDEWEB)

    Martino, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-08-08

    An evaluation of the previous Chemical Processing Cell (CPC) testing was performed to determine whether the planned concurrent operation, or “coupled” operations, of the Defense Waste Processing Facility (DWPF) with the Salt Waste Processing Facility (SWPF) has been adequately covered. Tests with the nitricglycolic acid flowsheet, which were both coupled and uncoupled with salt waste streams, included several tests that required extended boiling times. This report provides the evaluation of previous testing and the testing recommendation requested by Savannah River Remediation. The focus of the evaluation was impact on flammability in CPC vessels (i.e., hydrogen generation rate, SWPF solvent components, antifoam degradation products) and processing impacts (i.e., acid window, melter feed target, rheological properties, antifoam requirements, and chemical composition).

  4. Efficacy of backfilling and other engineered barriers in a radioactive waste repository in salt

    International Nuclear Information System (INIS)

    Claiborne, H.C.

    1982-09-01

    In the United States, investigation of potential host geologic formations was expanded in 1975 to include hard rocks. Potential groundwater intrusion is leading to very conservative and expensive waste package designs. Recent studies have concluded that incentives for engineered barriers and 1000-year canisters probably do not exist for reasonable breach scenarios. The assumption that multibarriers will significantly increase the safety margin is also questioned. Use of a bentonite backfill for surrounding a canister of exotic materials was developed in Sweden and is being considered in the US. The expectation that bentonite will remain essentially unchanged for hundreds of years for US repository designs may be unrealistic. In addition, thick bentonite backfills will increase the canister surface temperature and add much more water around the canister. The use of desiccant materials, such as CaO or MgO, for backfilling seems to be a better method of protecting the canister. An argument can also be made for not using backfill material in salt repositories since the 30-cm-thick space will provide for hole closure for many years and will promote heat transfer via natural convection. It is concluded that expensive safety systems are being considered for repository designs that do not necessarily increase the safety margin. It is recommended that the safety systems for waste repositories in different geologic media be addressed individually and that cost-benefit analyses be performed

  5. Mineral carbonation of phosphogypsum waste for production of useful carbonate and sulfate salts

    Directory of Open Access Journals (Sweden)

    Hannu-Petteri eMattila

    2015-11-01

    Full Text Available Phosphogypsum (CaSO4·2H2O waste is produced in large amounts during phosphoric acid (H3PO4 production. Minor quantities are utilized in construction or agriculture, while most of the material is stockpiled, creating an environmental challenge to prevent pollution of natural waters. In principle, the gypsum waste could be used to capture several hundred Mt of carbon dioxide (CO2. For example, when gypsum is converted to ammonium sulfate ((NH42SO4 with ammonia (NH3 and CO2, also solid calcium carbonate (CaCO3 is generated. The ammonium sulfate can be utilized as a fertilizer or in other mineral carbonation processes that use magnesium silicate-based rock as feedstock, while calcium carbonate has various uses as e.g. filler material. The reaction extent of the described process was studied by thermodynamic modeling and experimentally as a function of reactant concentrations and temperature. Other essential properties such as purity and quality of the solid products are also followed. Conversion efficiencies of >95% calcium from phosphogypsum to calcium carbonate are obtained. Scalenohedral, rhombohedral and prismatic calcite particles can be produced, though the precipitates contain certain contaminants such as rare earth metals and sulfur from the gypsum. A reverse osmosis membrane cartridge is also tested as an alternative and energy-efficient method of concentrating the ammonium sulfate salt solution instead of the traditional evaporation of the process solution.

  6. Experimental storage of high-level radioactive wastes in the Asse salt mine - technical aspects

    International Nuclear Information System (INIS)

    Gies, H.; Rothfuchs, T.; Feddersen, H.; Graefe, V.; Gross, S.; Hente, B.; Jockwer, N.; Kessels, W.; Schwaegermann, H.

    1988-01-01

    The work performed under this project in the Asse salt mine is an important milepost within the framework schedule of the 'Gorleben Poject' of Physikalisch-Technische Bundesanstalt (PTB). The project phase I (1982 - June 30, 1985) is about to be concluded at the time this report is published. The main points of interest of this project phase cover the planning of the experimental work, the design of experiments, and the first activities for developing the systems for handling the high-level radioactive wastes. The engineering development work has been advanced to the point where construction and manufacture of equipment can be started (transport containers Asse, TB1, collective transport containers, borehole gates, transport vehicles, waste positioning equipment, and borehole casing). Testing of the pipes for the last mentioned task with regard to the material's deformation behaviour will be done by the Dutch ECN as a sub-contractor. First laboratory experiments have been carried out on radiolysis gas formation, to complement the engineering work and the in-situ measuring programmes. (orig./RB) [de

  7. Sandia studies of high-level waste canisters and overpacks applicable for a salt repository

    International Nuclear Information System (INIS)

    Molecke, M.A.; Schaefer, D.W.; Glass, R.S.; Ruppen, J.A.

    1982-01-01

    An experimental program to develop candidate materials for use as high-level waste (HLW) overpacks or canisters in a salt repository has been in progress at Sandia National Laboratories since 1976. The main objective of this program is to provide a waste package barrier having a long lifetime in the chemical and physical environment of a repository. This paper summarizes the recent corrosion and metallurgical study results for the prime overpack material, TiCode-12, in the areas of uniform corrosion (extremely low rate and extent); local attack, e.g., pits and crevices (none were found); stress corrosion cracking susceptibility (no significant changes in macroscopic tensile properties were detected); hydrogen sorption-embrittlement effects; effects of gamma irradiation in solution; and sensitization effects (testing is still in process in the last three areas). Previous candidate screening analyses on other alloys and recent work on alternate overpack alloys are reviewed. All phases of these interrelated laboratory, hot-cell, and field experimental studies are described. 16 references, 8 figures, 4 tables

  8. SAVANNAH RIVER SITE INCIPIENT SLUDGE MIXING IN RADIOACTIVE LIQUID WASTE STORAGE TANKS DURING SALT SOLUTION BLENDING

    Energy Technology Data Exchange (ETDEWEB)

    Leishear, R.; Poirier, M.; Lee, S.; Steeper, T.; Fowley, M.; Parkinson, K.

    2011-01-12

    This paper is the second in a series of four publications to document ongoing pilot scale testing and computational fluid dynamics (CFD) modeling of mixing processes in 85 foot diameter, 1.3 million gallon, radioactive liquid waste, storage tanks at Savannah River Site (SRS). Homogeneous blending of salt solutions is required in waste tanks. Settled solids (i.e., sludge) are required to remain undisturbed on the bottom of waste tanks during blending. Suspension of sludge during blending may potentially release radiolytically generated hydrogen trapped in the sludge, which is a safety concern. The first paper (Leishear, et. al. [1]) presented pilot scale blending experiments of miscible fluids to provide initial design requirements for a full scale blending pump. Scaling techniques for an 8 foot diameter pilot scale tank were also justified in that work. This second paper describes the overall reasons to perform tests, and documents pilot scale experiments performed to investigate disturbance of sludge, using non-radioactive sludge simulants. A third paper will document pilot scale CFD modeling for comparison to experimental pilot scale test results for both blending tests and sludge disturbance tests. That paper will also describe full scale CFD results. The final paper will document additional blending test results for stratified layers in salt solutions, scale up techniques, final full scale pump design recommendations, and operational recommendations. Specifically, this paper documents a series of pilot scale tests, where sludge simulant disturbance due to a blending pump or transfer pump are investigated. A principle design requirement for a blending pump is UoD, where Uo is the pump discharge nozzle velocity, and D is the nozzle diameter. Pilot scale test results showed that sludge was undisturbed below UoD = 0.47 ft{sup 2}/s, and that below UoD = 0.58 ft{sup 2}/s minimal sludge disturbance was observed. If sludge is minimally disturbed, hydrogen will not be

  9. Renal Dysfunction Induced by Kidney-Specific Gene Deletion of Hsd11b2 as a Primary Cause of Salt-Dependent Hypertension.

    Science.gov (United States)

    Ueda, Kohei; Nishimoto, Mitsuhiro; Hirohama, Daigoro; Ayuzawa, Nobuhiro; Kawarazaki, Wakako; Watanabe, Atsushi; Shimosawa, Tatsuo; Loffing, Johannes; Zhang, Ming-Zhi; Marumo, Takeshi; Fujita, Toshiro

    2017-07-01

    Genome-wide analysis of renal sodium-transporting system has identified specific variations of Mendelian hypertensive disorders, including HSD11B2 gene variants in apparent mineralocorticoid excess. However, these genetic variations in extrarenal tissue can be involved in developing hypertension, as demonstrated in former studies using global and brain-specific Hsd11b2 knockout rodents. To re-examine the importance of renal dysfunction on developing hypertension, we generated kidney-specific Hsd11b2 knockout mice. The knockout mice exhibited systemic hypertension, which was abolished by reducing salt intake, suggesting its salt-dependency. In addition, we detected an increase in renal membrane expressions of cleaved epithelial sodium channel-α and T53-phosphorylated Na + -Cl - cotransporter in the knockout mice. Acute intraperitoneal administration of amiloride-induced natriuresis and increased urinary sodium/potassium ratio more in the knockout mice compared with those in the wild-type control mice. Chronic administration of amiloride and high-KCl diet significantly decreased mean blood pressure in the knockout mice, which was accompanied with the correction of hypokalemia and the resultant decrease in Na + -Cl - cotransporter phosphorylation. Accordingly, a Na + -Cl - cotransporter blocker hydrochlorothiazide significantly decreased mean blood pressure in the knockout mice. Chronic administration of mineralocorticoid receptor antagonist spironolactone significantly decreased mean blood pressure of the knockout mice along with downregulation of cleaved epithelial sodium channel-α and phosphorylated Na + -Cl - cotransporter expression in the knockout kidney. Our data suggest that kidney-specific deficiency of 11β-HSD2 leads to salt-dependent hypertension, which is attributed to mineralocorticoid receptor-epithelial sodium channel-Na + -Cl - cotransporter activation in the kidney, and provides evidence that renal dysfunction is essential for developing the

  10. A reactive distillation process for the treatment of LiCl-KCl eutectic waste salt containing rare earth chlorides

    Energy Technology Data Exchange (ETDEWEB)

    Eun, H.C., E-mail: ehc2004@kaeri.re.kr; Choi, J.H.; Kim, N.Y.; Lee, T.K.; Han, S.Y.; Lee, K.R.; Park, H.S.; Ahn, D.H.

    2016-11-15

    The pyrochemical process, which recovers useful resources (U/TRU metals) from used nuclear fuel using an electrochemical method, generates LiCl-KCl eutectic waste salt containing radioactive rare earth chlorides (RECl{sub 3}). It is necessary to develop a simple process for the treatment of LiCl-KCl eutectic waste salt in a hot-cell facility. For this reason, a reactive distillation process using a chemical agent was achieved as a method to separate rare earths from the LiCl-KCl waste salt. Before conducting the reactive distillation, thermodynamic equilibrium behaviors of the reactions between rare earth (Nd, La, Ce, Pr) chlorides and the chemical agent (K{sub 2}CO{sub 3}) were predicted using software. The addition of the chemical agent was determined to separate the rare earth chlorides into an oxide form using these equilibrium results. In the reactive distillation test, the rare earth chlorides in LiCl-KCl eutectic salt were decontaminated at a decontamination factor (DF) of more than 5000, and were mainly converted into oxide (Nd{sub 2}O{sub 3}, CeO{sub 2}, La{sub 2}O{sub 3}, Pr{sub 2}O{sub 3}) or oxychloride (LaOCl, PrOCl) forms. The LiCl-KCl was purified into a form with a very low concentration (<1 ppm) for the rare earth chlorides.

  11. Temperature calculations on different configurations for disposal of high-level reprocessing waste in a salt dome model

    International Nuclear Information System (INIS)

    Hamstra, J.; Kevenaar, J.W.A.M.

    1978-06-01

    A medium size salt dome is considered as a structure in which a repository can be located for all radioactive wastes to be produced within the scope of a postulated nuclear power program. A dominating design factor for the lay-out of such a waste repository is the temperature distribution in the salt dome resulting from decay heat released from the buried solidified high-level reprocessing waste. Two numerical models are presented for the calculation of both global and local rock salt temperatures. The results of calculations performed with these models are demonstrated to be compatible. Rock salt temperatures related to several types of burial configurations, ranging from two layer configurations with various vertical distances between the layers via a three and a four layer repository to deep bore hole concepts varying from 100 to 600 m bore hole depth, can therefore be calculated with one rather simple unit cell model. The results of these calculations indicate that rock salt temperatures can be kept within acceptable limits to realize a repository using standard mining techniques. The temperatures at mine galery level prove to be a dominating factor in the selection of a repository configuration. More detailed calculations of these temperatures taking into account the loading sequence and the cooling capacity of the mine ventilation are recommended. Finally the apparent advantages of a deep bore hole concept emphasize the need for R and D work with respect to advanced drilling techniques in order to achieve deep dry disposal bore holes that can be realized from a burial mine in the salt dome. (Auth.)

  12. Mineral sources of water and their influence on the safe disposal of radioactive wastes in bedded salt deposits

    Energy Technology Data Exchange (ETDEWEB)

    Fallis, S.M.

    1973-12-01

    With the increased use of nuclear energy, there will be subsequent increases in high-level radioactive wastes such as Sr/sup 90/, Cs/sup 137/, and Pu/sup 239/. Several agencies have considered the safest possible means to store or dispose of wastes in geologic environments such as underground storage in salt deposits, shale beds, abandoned dry mines, and in clay and shale pits. Salt deposits have received the most favorable attention because they exist in dry environments and because of other desirable properties of halite (its plasticity, gamma-ray shielding, heat dissipation ability, low mining cost, and worldwide abundance). Much work has been done on bedded salt deposits, particularly the Hutchinson Salt Member of the Wellington Formation at Lyons, Kansas. Salt beds heated by the decay of the radioactive wastes may release water by dehydration of hydrous minerals commonly present in evaporite sequences or water present in other forms such as fluid inclusions. More than 80 hydrous minerals are known to occur in evaporite deposits. The occurrences, total water contents (up to 63%) and dehydration temperatures (often less that 150/sup 0/C) of these minerals are given. Since it is desirable to dispose of radioactive wastes in a dry environment, care must be taken that large quantities of water are not released through the heating of hydrous minerals. Seventy-four samples from four cores taken at Lyons, Kansas, were analyzed by x-ray diffraction. The minerals detected were halite, anhydrite, gypsum, polyhalite, dolomite, magnesite, quartz, feldspar, and the clay minerals illite, chlorite, kaolinite, vermiculite, smectite, mixed-layer clay, and corrensite (interstratified chlorite-vermiculite). Of these, gypsum, polyhalite and the clay minerals are all capable of releasing water when heated.

  13. Mineral sources of water and their influence on the safe disposal of radioactive wastes in bedded salt deposits

    International Nuclear Information System (INIS)

    Fallis, S.M.

    1973-12-01

    With the increased use of nuclear energy, there will be subsequent increases in high-level radioactive wastes such as Sr 90 , Cs 137 , and Pu 239 . Several agencies have considered the safest possible means to store or dispose of wastes in geologic environments such as underground storage in salt deposits, shale beds, abandoned dry mines, and in clay and shale pits. Salt deposits have received the most favorable attention because they exist in dry environments and because of other desirable properties of halite (its plasticity, gamma-ray shielding, heat dissipation ability, low mining cost, and worldwide abundance). Much work has been done on bedded salt deposits, particularly the Hutchinson Salt Member of the Wellington Formation at Lyons, Kansas. Salt beds heated by the decay of the radioactive wastes may release water by dehydration of hydrous minerals commonly present in evaporite sequences or water present in other forms such as fluid inclusions. More than 80 hydrous minerals are known to occur in evaporite deposits. The occurrences, total water contents (up to 63%) and dehydration temperatures (often less that 150 0 C) of these minerals are given. Since it is desirable to dispose of radioactive wastes in a dry environment, care must be taken that large quantities of water are not released through the heating of hydrous minerals. Seventy-four samples from four cores taken at Lyons, Kansas, were analyzed by x-ray diffraction. The minerals detected were halite, anhydrite, gypsum, polyhalite, dolomite, magnesite, quartz, feldspar, and the clay minerals illite, chlorite, kaolinite, vermiculite, smectite, mixed-layer clay, and corrensite (interstratified chlorite-vermiculite). Of these, gypsum, polyhalite and the clay minerals are all capable of releasing water when heated

  14. Radioactive waste isolation in salt: special advisory report on the status of the Office of Nuclear Waste Isolation's plans for repository performance assessment

    International Nuclear Information System (INIS)

    Ditmars, J.D.; Walbridge, E.W.; Rote, D.M.; Harrison, W.; Herzenberg, C.L.

    1983-10-01

    Repository performance assessment is analysis that identifies events and processes that might affect a repository system for isolation of radioactive waste, examines their effects on barriers to waste migration, and estimates the probabilities of their occurrence and their consequences. In 1983 Battelle Memorial Institute's Office of Nuclear Waste Isolation (ONWI) prepared two plans - one for performance assessment for a waste repository in salt and one for verification and validation of performance assessment technology. At the request of the US Department of Energy's Salt Repository Project Office (SRPO), Argonne National Laboratory reviewed those plans and prepared this report to advise SRPO of specific areas where ONWI's plans for performance assessment might be improved. This report presents a framework for repository performance assessment that clearly identifies the relationships among the disposal problems, the processes underlying the problems, the tools for assessment (computer codes), and the data. In particular, the relationships among important processes and 26 model codes available to ONWI are indicated. A common suggestion for computer code verification and validation is the need for specific and unambiguous documentation of the results of performance assessment activities. A major portion of this report consists of status summaries of 27 model codes indicated as potentially useful by ONWI. The code summaries focus on three main areas: (1) the code's purpose, capabilities, and limitations; (2) status of the elements of documentation and review essential for code verification and validation; and (3) proposed application of the code for performance assessment of salt repository systems. 15 references, 6 figures, 4 tables

  15. The use of marine aquaculture solid waste for nursery production of the salt marsh plants Spartina alterniflora and Juncus roemerianus

    Directory of Open Access Journals (Sweden)

    H.M. Joesting

    2016-05-01

    Full Text Available Recent technological advances in marine shrimp and finfish aquaculture alleviate many of the environmental risks associated with traditional aquaculture, but challenges remain in cost-effective waste management. Liquid effluent from freshwater aquaculture systems has been shown to be effective in agricultural crop production (i.e., aquaponics, but few studies have explored the potential for reuse of marine aquaculture effluent, particularly the solid fraction. The purpose of this study was to investigate the use of marine aquaculture solid waste as a nutrient source for the nursery production of two salt tolerant plants commonly used in coastal salt marsh restoration, Spartina alterniflora (smooth cordgrass and Juncus roemerianus (black needlerush. Specifically, measurements of plant biomass and tissue nitrogen and phosphorus allocation were compared between plants fertilized with dried shrimp biofloc solids and unfertilized controls, as well as between plants fertilized with dried fish solids and unfertilized controls. In both experiments, S. alterniflora plants fertilized with marine aquaculture solids showed few significant differences from unfertilized controls, whereas fertilized J. roemerianus plants had significantly greater biomass and absorbed and incorporated more nutrients in plant tissue compared to unfertilized controls. These results suggest that J. roemerianus may be a suitable plant species for the remediation of marine aquaculture solid waste. Keywords: Marine aquaculture, Salt marsh plants, Solid waste, Phytoremediation

  16. Disposal of high-level waste from nuclear power plants in Denmark. Salt dome investigations. v.4

    International Nuclear Information System (INIS)

    1981-01-01

    The present report deals with construction, operation and sealing of disposal facilities for high-level waste in a salt dome. It is volume 4 of five volumes that together constitute the final report on the Danish utilities' salt dome investigations. The safety investigations were carried out for a deep-hole disposal facility located in the salt dome on Mors. In principle the results of the investigations also apply to a shaft/mine disposal facility. The facility is designed for the disposal of vitrified high-level waste in the shape of glass canisters. There is a low concentration of waste in each canister, approx. 10%. Furthermore, it was selected to place the waste in an intermediate storage for about 40 years prior to its final disposal. Consequently, heat generation in the waste at the time of final disposal will be modest, resulting in low temperature increase in the salt. As an example, the highest temperature increase will be approx. 40 deg. C. and it will occur at the edge of the hole five years after disposal has taken place. Prior to disposal, the glass canisters are encased in steel casks with 15 cm thick walls. Three canisters are placed in each cask, and 215 casks are stacked on top on one another in each deep-hole from a depth of 1200 m to 2500 m underground. The additional encasing is designed to protect the glass from dissolution should any brine reach the disposal facility. Furthermore, the steel cask protects the glass canisters against pressure from the wall of the hole. The technical design of the disposal facility gives it a considerable safety margin against unexpected events. The investigations proved Cretaceous strata to constitute an effective secondary barrier that would prevent radioactive matter from travelling from the underlying disposal facility to the biosphere. (BP)

  17. A thermodynamic approach on vapor-condensation of corrosive salts from flue gas on boiler tubes in waste incinerators

    International Nuclear Information System (INIS)

    Otsuka, Nobuo

    2008-01-01

    Thermodynamic equilibrium calculation was conducted to understand the effects of tube wall temperature, flue gas temperature, and waste chemistry on the type and amount of vapor-condensed 'corrosive' salts from flue gas on superheater and waterwall tubes in waste incinerators. The amount of vapor-condensed compounds from flue gases at 650-950 deg. C on tube walls at 350-850 deg. C was calculated, upon combustion of 100 g waste with 1.6 stoichiometry (in terms of the air-fuel ratio). Flue gas temperature, rather than tube wall temperature, influenced the deposit chemistry of boiler tubes significantly. Chlorine, sulfur, sodium, potassium, and calcium contents in waste affected it as well

  18. Costs for off-site disposal of nonhazardous oil field wastes: Salt caverns versus other disposal methods

    Energy Technology Data Exchange (ETDEWEB)

    Veil, J.A.

    1997-09-01

    According to an American Petroleum Institute production waste survey reported on by P.G. Wakim in 1987 and 1988, the exploration and production segment of the US oil and gas industry generated more than 360 million barrels (bbl) of drilling wastes, more than 20 billion bbl of produced water, and nearly 12 million bbl of associated wastes in 1985. Current exploration and production activities are believed to be generating comparable quantities of these oil field wastes. Wakim estimates that 28% of drilling wastes, less than 2% of produced water, and 52% of associated wastes are disposed of in off-site commercial facilities. In recent years, interest in disposing of oil field wastes in solution-mined salt caverns has been growing. This report provides information on the availability of commercial disposal companies in oil-and gas-producing states, the treatment and disposal methods they employ, and the amounts they charge. It also compares cavern disposal costs with the costs of other forms of waste disposal.

  19. Mineral Carbonation of Phosphogypsum Waste for Production of Useful Carbonate and Sulfate Salts

    Energy Technology Data Exchange (ETDEWEB)

    Mattila, Hannu-Petteri, E-mail: hmattila@abo.fi; Zevenhoven, Ron [Thermal and Flow Engineering Laboratory, Åbo Akademi University, Turku (Finland)

    2015-11-16

    Phosphogypsum (CaSO{sub 4}·2H{sub 2}O, PG) waste is produced in large amounts during phosphoric acid (H{sub 3}PO{sub 4}) production. Minor quantities are utilized in construction or agriculture, while most of the material is stockpiled, creating an environmental challenge to prevent pollution of natural waters. In principle, the gypsum waste could be used to capture several hundred megatonnes of carbon dioxide (CO{sub 2}). For example, when gypsum is converted to ammonium sulfate [(NH{sub 4}){sub 2}SO{sub 4}] with ammonia (NH{sub 3}) and CO{sub 2}, also solid calcium carbonate (CaCO{sub 3}) is generated. The ammonium sulfate can be utilized as a fertilizer or in other mineral carbonation processes that use magnesium silicate-based rock as feedstock, while calcium carbonate has various uses as, e.g., filler material. The reaction extent of the described process was studied by thermodynamic modeling and experimentally as a function of reactant concentrations and temperature. Other essential properties such as purity and quality of the solid products are also followed. Conversion efficiencies of >95% calcium from PG to calcium carbonate are obtained. Scalenohedral, rhombohedral, and prismatic calcite particles can be produced, although the precipitates contain certain contaminants such as rare earth metals and sulfur from the gypsum. A reverse osmosis membrane cartridge is also tested as an alternative and energy-efficient method of concentrating the ammonium sulfate salt solution instead of the traditional evaporation of the process solution.

  20. Radioactive waste isolation in salt: Peer review of the Office of Nuclear Waste Isolation's draft report on an issues hierarchy and data needs for site characterization

    International Nuclear Information System (INIS)

    Harrison, W.; Fenster, D.F.; Ditmars, J.D.; Paddock, R.A.; Rote, D.M.; Hambley, D.F.; Seitz, M.G.; Hull, A.B.

    1986-12-01

    At the request of the Salt Repository Project (SRPO), Argonne National Laboratory conducted an independent peer review of a report by the Battelle Office of Nuclear Waste Isolation entitled ''Salt Repository Project Issues Hierarchy and Data Needs for Site Characterization (Draft).'' This report provided a logical structure for evaluating the outstanding questions (issues) related to selection and licensing of a site as a high-level waste repository. It also provided a first estimate of the information and data necessary to answer or resolve those questions. As such, this report is the first step in developing a strategy for site characterization. Microfiche copies of ''Draft Issues Hierarchy, Resolution Strategy, and Information Needs for Site Characterization and Environmental/Socioeconomic Evaluation - July, 1986'' and ''Issues Hierarchy and Data Needs for Site Characterization - February, 1985'' are included in the back pocket of this report

  1. Radioactive waste isolation in salt: peer review of Westinghouse Electric Corporation's report on reference conceptual designs for a repository waste package

    International Nuclear Information System (INIS)

    Rote, D.M.; Hull, A.B.; Was, G.S.; Macdonald, D.D.; Wilde, B.E.; Russell, J.E.; Kruger, J.; Harrison, W.; Hambley, D.F.

    1985-10-01

    This report documents the findings of the peer panel constituted by Argonne National Laboratory to review Region A of Westinghouse Electric Corporation's report entitled Waste Package Reference Conceptual Designs for a Repository in Salt. The panel determined that the reviewed report does not provide reasonable assurance that US Nuclear Regulatory Commission (NRC) requirements for waste packages will be met by the proposed design. It also found that it is premature to call the design a ''reference design,'' or even a ''reference conceptual design.'' This review report provides guidance for the preparation of a more acceptable design document

  2. Radioactive waste isolation in salt: peer review of Westinghouse Electric Corporation's report on reference conceptual designs for a repository waste package

    Energy Technology Data Exchange (ETDEWEB)

    Rote, D.M.; Hull, A.B.; Was, G.S.; Macdonald, D.D.; Wilde, B.E.; Russell, J.E.; Kruger, J.; Harrison, W.; Hambley, D.F.

    1985-10-01

    This report documents the findings of the peer panel constituted by Argonne National Laboratory to review Region A of Westinghouse Electric Corporation's report entitled Waste Package Reference Conceptual Designs for a Repository in Salt. The panel determined that the reviewed report does not provide reasonable assurance that US Nuclear Regulatory Commission (NRC) requirements for waste packages will be met by the proposed design. It also found that it is premature to call the design a ''reference design,'' or even a ''reference conceptual design.'' This review report provides guidance for the preparation of a more acceptable design document.

  3. Extraction, Scrub, and Strip Test Results for the Salt Waste Processing Facility Caustic Side Solvent Extraction Solvent Sample

    Energy Technology Data Exchange (ETDEWEB)

    Peters, T. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-10-06

    An Extraction, Scrub, and Strip (ESS) test was performed on a sample of Salt Waste Processing Facility (SWPF) Caustic-Side Solvent Extraction (CSSX) solvent and salt simulant to determine cesium distribution ratios (D(Cs)), and cesium concentration in the strip effluent (SE) and decontaminated salt solution (DSS) streams; this data will be used by Parsons to help determine if the solvent is qualified for use at the SWPF. The ESS test showed acceptable performance of the solvent for extraction, scrub, and strip operations. The extraction D(Cs) measured 12.5, exceeding the required value of 8. This value is consistent with results from previous ESS tests using similar solvent formulations. Similarly, scrub and strip cesium distribution ratios fell within acceptable ranges. This revision was created to correct an error. The previous revision used an incorrect set of temperature correction coefficients which resulted in slight deviations from the correct D(Cs) results.

  4. Extraction, scrub, and strip test results for the salt waste processing facility caustic side solvent extraction solvent example

    Energy Technology Data Exchange (ETDEWEB)

    Peters, T. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-08-01

    An Extraction, Scrub, and Strip (ESS) test was performed on a sample of Salt Waste Processing Facility (SWPF) Caustic-Side Solvent Extraction (CSSX) solvent and salt simulant to determine cesium distribution ratios (D(Cs)), and cesium concentration in the strip effluent (SE) and decontaminated salt solution (DSS) streams; this data will be used by Parsons to help determine if the solvent is qualified for use at the SWPF. The ESS test showed acceptable performance of the solvent for extraction, scrub, and strip operations. The extraction D(Cs) measured 12.9, exceeding the required value of 8. This value is consistent with results from previous ESS tests using similar solvent formulations. Similarly, scrub and strip cesium distribution ratios fell within acceptable ranges.

  5. Resistance of Coatings for Boiler Components of Waste-to-Energy Plants to Salt Melts Containing Copper Compounds

    Science.gov (United States)

    Galetz, Mathias Christian; Bauer, Johannes Thomas; Schütze, Michael; Noguchi, Manabu; Cho, Hiromitsu

    2013-06-01

    The accelerating effect of heavy metal compounds on the corrosive attack of boiler components like superheaters poses a severe problem in modern waste-to-energy plants (WTPs). Coatings are a possible solution to protect cheap, low alloyed steel substrates from heavy metal chloride and sulfate salts, which have a relatively low melting point. These salts dissolve many alloys, and therefore often are the limiting factor as far as the lifetime of superheater tubes is concerned. In this work the corrosion performance under artificial salt deposits of different coatings, manufactured by overlay welding, thermal spraying of self-fluxing as well as conventional systems was investigated. The results of our studies clearly demonstrate the importance of alloying elements such as molybdenum or silicon. Additionally, the coatings have to be dense and of a certain thickness in order to resist the corrosive attack under these severe conditions.

  6. The sodium-bicarbonate cotransporter NBCe2 (slc4a5) expressed in human renal proximal tubules shows increased apical expression under high-salt conditions.

    Science.gov (United States)

    Gildea, John J; Xu, Peng; Carlson, Julia M; Gaglione, Robert T; Bigler Wang, Dora; Kemp, Brandon A; Reyes, Camellia M; McGrath, Helen E; Carey, Robert M; Jose, Pedro A; Felder, Robin A

    2015-12-01

    The electrogenic sodium bicarbonate cotransporter (NBCe2) is encoded by SLC4A5, variants of which have been associated with salt sensitivity of blood pressure, which affects 25% of the adult population. NBCe2 is thought to mediate sodium bicarbonate cotransport primarily in the renal collecting duct, but NBCe2 mRNA is also found in the rodent renal proximal tubule (RPT). The protein expression or function of NBCe2 has not been demonstrated in the human RPT. We validated an NBCe2 antibody by shRNA and Western blot analysis, as well as overexpression of an epitope-tagged NBCe2 construct in both RPT cells (RPTCs) and human embryonic kidney 293 (HEK293) cells. Using this validated NBCe2 antibody, we found NBCe2 protein expression in the RPT of fresh and frozen human kidney slices, RPTCs isolated from human urine, and isolated RPTC apical membrane. Under basal conditions, NBCe2 was primarily found in the Golgi, while NBCe1 was primarily found at the basolateral membrane. Following an acute short-term increase in intracellular sodium, NBCe2 expression was increased at the apical membrane in cultured slices of human kidney and polarized, immortalized RPTCs. Sodium bicarbonate transport was increased by monensin and overexpression of NBCe2, decreased by NBCe2 shRNA, but not by NBCe1 shRNA, and blocked by 2,2'-(1,2-ethenediyl)bis[5-isothiocyanato-benzenesulfonic acid]. NBCe2 could be important in apical sodium and bicarbonate cotransport under high-salt conditions; the implication of the ex vivo studies to the in vivo situation when salt intake is increased remains unclear. Therefore, future studies will examine the role of NBCe2 in mediating increased renal sodium transport in humans whose blood pressures are elevated by an increase in sodium intake. Copyright © 2015 the American Physiological Society.

  7. Surface displacements and pillar stresses associated with nuclear waste disposal in salt

    International Nuclear Information System (INIS)

    Hardy, M.P.; St John, C.M.

    1977-01-01

    A numerical model for regional analysis of stresses and displacement, resulting from heat generating waste placement in underground salt excavations, is presented. The model, which is an extension of that described by McClain and Starfield (1971), is based upon the displacement discontinuity method of stress analysis. It incorporates an empirical characterization of creep behavior of material on the excavation horizon and accounts for thermally induced stresses and displacements. The versatility of this approach is illustrated by the results of three relatively short simulations of test scale disposal facilities at shallow and greater depths. In addition, a three-dimensional code was used to evaluate the surface displacement history for a full-scale repository. This latter code, a thermoelastic analysis, gives an upper bound for the surface movements. It is concluded that the pillar stresses are the result of a complex non-linear interaction of many variables, and the maximum pillar stress can reach several multiples of the tributory-area pillar stress

  8. Brine: a computer program to compute brine migration adjacent to a nuclear waste canister in a salt repository

    International Nuclear Information System (INIS)

    Duckworth, G.D.; Fuller, M.E.

    1980-01-01

    This report presents a mathematical model used to predict brine migration toward a nuclear waste canister in a bedded salt repository. The mathematical model is implemented in a computer program called BRINE. The program is written in FORTRAN and executes in the batch mode on a CDC 7600. A description of the program input requirements and output available is included. Samples of input and output are given

  9. Safety evaluation of geological disposal concepts for low and medium-level wastes in rock-salt (Pacoma project)

    International Nuclear Information System (INIS)

    Prij, J.; Van Dalen, A.; Roodbergen, H.A.; Slagter, W.; Van Weers, A.W.; Zanstra, D.A.; Glasbergen, P.; Koester, H.W.; Lembrechts, J.F.; Nijhof-Pan, I.; Slot, A.F.M.

    1991-01-01

    In the framework of the Performance Assessment of Confinements for MLW and Alpha Waste (PACOMA) the disposal options dealing with rock-salt are studied by GSF and ECN (with subcontract to RIVM). The overall objectives of these studies are to develop and demonstrate procedures for the radiological safety assessment of a deep repository in salt formations. An essential objective is to show how far appropriate choices of the repository design parameters can improve the performances of the whole system. The research covers two waste inventories (the Dutch OPLA and the PACOMA reference inventory), two disposal techniques (conventional and solution mining) and three types of formations (salt dome, pillow and bedded salt). An important part of the research has been carried out in the socalled VEOS project within the framework of the Dutch OPLA study. The methodology used in the consequence analysis is a deterministic one. The models and calculation tools used to perform the consequence analysis are the codes: EMOS, METROPOL and BIOS. The results are expressed in terms of dose rates and doses to individuals as well as to groups. Detailed information with respect to the input data and the results obtained with the three codes is given in three annexes to this final report

  10. Geosphere migration studies as support for the comparison of candidate sites for disposal of radioactive waste in rock-salt

    International Nuclear Information System (INIS)

    Glasbergen, P.; Hassanizadeh, S.M.; Noordijk, H.; Sauter, F.

    1988-01-01

    The Dutch research program on the geological disposal of radioactive waste was designed to supply a basis for the selection of combinations of three factors, i.e., type of rock-salt formation, site, and disposal technique, satisfying radiological standards and other criteria for final disposal. The potential sites have been grouped according to the type of rock-salt formation (e.g. bedded salt and salt domes) and two classes of depth below the surface of the ground. Values for geohydrological parameters were obtained by extrapolation of data from existing boreholes and analysis of the sedimentary environment. A three-dimensional model of groundwater flow and contaminant transport, called METROPOL, has been developed. To investigate the effect of high salinity on nuclide transport properly, a theoretical experimental study was carried out. Use of a thermodynamic approach showed that terms related to salt mass fraction have to be added to Darcy's and Fick's laws. An experimental study to investigate effects of these modifications is in progress. 8 refs.; 8 figs.; 1 table

  11. Disposal of high-level waste from nuclear power plants in Denmark. Salt dome investigations. v.2

    International Nuclear Information System (INIS)

    1981-01-01

    The present report deals with the geological investigations performed for determing the feasibility of a repository for high-level waste in a salt dome. It is volume 2 of five volumes that together constitute the final report of the Danish utilities' salt dome investigations. The purpose of the work was to procure a more detailed knowledge of the geology of salt domes in North Jutland on example of Mors. The Mors dome is oval with the two axes of approx. 12.5 km and 8 km respectively. Two deep wells have been drilled into the salt. These wells reach 3400-3500 m below surface. Until a depth of about 3200 m Erslev 2 passes through rock salt of Zechstein 1 which is the oldest evaporite series. However, it could also be interlayed with the slightly younger Zechstein 2. At about 3200 m a marker layer was met with Zechstein 2 salt below. Interpretation of cores and results of downhole electromagnetic and borehole gravimetric measurements show that there is a large area around Erslev 2 which consists of very pure sodium chloride with traces of anhydrite (calcium, sulphate) 1-3%. This area is used for the repository design and safety evaluation. The hydrological conditions existing in the strata above the salt dome (caprock) have been investigated with the help of four hydrogeological wells, placed two each, on two different sites. The cores themselves were taken at various depths in all four holes. With these laboratory methods it has been possible to measure data relevant to hydrology - such as porosity and permeability - as well as geochemistry. (BP)

  12. Distillation Separation of Hydrofluoric Acid and Nitric Acid from Acid Waste Using the Salt Effect on Vapor-Liquid Equilibrium

    Science.gov (United States)

    Yamamoto, Hideki; Sumoge, Iwao

    2011-03-01

    This study presents the distillation separation of hydrofluoric acid with use of the salt effect on the vapor-liquid equilibrium for acid aqueous solutions and acid mixtures. The vapor-liquid equilibrium of hydrofluoric acid + salt systems (fluorite, potassium nitrate, cesium nitrate) was measured using an apparatus made of perfluoro alkylvinylether. Cesium nitrate showed a salting-out effect on the vapor-liquid equilibrium of the hydrofluoric acid-water system. Fluorite and potassium nitrate showed a salting-in effect on the hydrofluoric acid-water system. Separation of hydrofluoric acid from an acid mixture containing nitric acid and hydrofluoric acid was tested by the simple distillation treatment using the salt effect of cesium nitrate (45 mass%). An acid mixture of nitric acid (5.0 mol · dm-3) and hydrofluoric acid (5.0 mol · dm-3) was prepared as a sample solution for distillation tests. The concentration of nitric acid in the first distillate decreased from 5.0 mol · dm-3 to 1.13 mol · dm-3, and the concentration of hydrofluoric acid increased to 5.41 mol · dm-3. This first distillate was further distilled without the addition of salt. The concentrations of hydrofluoric acid and nitric acid in the second distillate were 7.21 mol · dm-3 and 0.46 mol · dm-3, respectively. It was thus found that the salt effect on vapor-liquid equilibrium of acid mixtures was effective for the recycling of acids from acid mixture wastes.

  13. HNF1B mutations associate with hypomagnesemia and renal magnesium wasting

    NARCIS (Netherlands)

    Adalat, Shazia; Woolf, Adrian S.; Johnstone, Karen A.; Wirsing, Andrea; Harries, Lorna W.; Long, David A.; Hennekam, Raoul C.; Ledermann, Sarah E.; Rees, Lesley; van't Hoff, William; Marks, Stephen D.; Trompeter, Richard S.; Tullus, Kjell; Winyard, Paul J.; Cansick, Janette; Mushtaq, Imran; Dhillon, Harjeeta K.; Bingham, Coralie; Edghill, Emma L.; Shroff, Rukshana; Stanescu, Horia; Ryffel, Gerhart U.; Ellard, Sian; Bockenhauer, Detlef

    2009-01-01

    Mutations in hepatocyte nuclear factor 1B (HNF1B), which is a transcription factor expressed in tissues including renal epithelia, associate with abnormal renal development. While studying renal phenotypes of children with HNF1B mutations, we identified a teenager who presented with tetany and

  14. Development of waste package designs for disposal in a salt repository

    International Nuclear Information System (INIS)

    Balmert, M.E.

    1983-01-01

    Three package design concepts were developed for CHLW and DHLW forms and spent fuel rods: (1) carbon steel overpack, borehole emplacement, (2) titanium clad, carbon steel reinforced overpack, borehole emplacement, and (3) carbon steel (self-shield) overpack, tunnel emplacement. For a DHLW canister with titanium clad overpack, the concept features a 9.5-cm-thick carbon steel overpack reinforcement supporting a 0.25-cm-thick titanium shell. The overall package dimensions are 84 cm diameter x 340 cm long weighing about 8.8 mtons. By contrast, a monolithic DHLW borehole package has a carbon steel overpack that is 10.4 cm thick, weighing about 9.3 mtons. The titanium clad/carbon steel reinforced borehole package is intended for remote emplacement in a vertical borehole in salt. The carbon steel overpack reinforcement provides structural integrity, primarily to resist external pressure, while the titanium overpack provides the necessary corrosion resistance to meet containment requirements. The carbon steel borehole package concept provides containment integrity for both external pressure and corrosion environments with a thicker carbon steel overpack in place of the titanium/carbon steel concept. A third concept utilizes an even greater thickness of cast steel or iron to resist external pressure and corrosion as well as reduce external shielding requirements. For example, a cast steel DHLW package would have overall dimensions of 125 cm diameter x 390 cm long, weighing 31 mtons. The purpose of this self-shield concept is to minimize handling and emplacement operations by reducing the package surface radiation dose to about 100 mrem/hr. In addition, it may serve as a shipping cask, thereby eliminating the need for a shielded hot cell at the repository for waste package assembly operations. 7 figures

  15. Treatment of radioactive waste salt by using synthetic silica-based phosphate composite for de-chlorination and solidification

    Science.gov (United States)

    Cho, In-Hak; Park, Hwan-Seo; Lee, Ki-Rak; Choi, Jung-Hun; Kim, In-Tae; Hur, Jin Mok; Lee, Young-Seak

    2017-09-01

    In the radioactive waste management, waste salts as metal chloride generated from a pyrochemical process to recover uranium and transuranic elements are one of problematic wastes due to their intrinsic properties such as high volatility and low compatibility with conventional glasses. This study reports a method to stabilize and solidify LiCl waste via de-chlorination using a synthetic composite, U-SAP (SiO2-Al2O3-B2O3-Fe2O3-P2O5) prepared by a sol-gel process. The composite was reacted with alkali metal elements to produce some metal aluminosilicates, aluminophosphates or orthophosphate as a crystalline or amorphous compound. Different from the original SAP (SiO2-Al2O3-P2O5), the reaction product of U-SAP could be successfully fabricated as a monolithic wasteform without a glassy binder at a proper reaction/consolidation condition. From the results of the FE-SEM, FT-IR and MAS-NMR analysis, it could be inferred that the Si-rich phase and P-rich phase as a glassy grains would be distributed in tens of nm scale, where alkali metal elements would be chemically interacted with Si-rich or P-rich region in the virgin U-SAP composite and its products was vitrified into a silicate or phosphate glass after a heat-treatment at 1150 °C. The PCT-A (Product Consistency Test, ASTM-1208) revealed that the mass loss of Cs and Sr in the U-SAP wasteform had a range of 10-3∼10-1 g/m2 and the leach-resistance of the U-SAP wasteform was comparable to other conventional wasteforms. From the U-SAP method, LiCl waste salt was effectively stabilized and solidified with high waste loading and good leach-resistance.

  16. Biochemical solubilization of toxic salts from residual geothermal brines and waste waters

    Science.gov (United States)

    Premuzic, Eugene T.; Lin, Mow S.

    1994-11-22

    A method of solubilizing metal salts such as metal sulfides in a geothermal sludge using mutant Thiobacilli selected for their ability to metabolize metal salts at high temperature is disclosed, The method includes the introduction of mutated Thiobacillus ferrooxidans and Thiobacillus thiooxidans to a geothermal sludge or brine. The microorganisms catalyze the solubilization of metal salts, For instance, in the case of metal sulfides, the microorganisms catalyze the solubilization to form soluble metal sulfates.

  17. Numerical analysis of impurity separation from waste salt by investigating the change of concentration at the interface during zone refining process

    Science.gov (United States)

    Choi, Ho-Gil; Shim, Moonsoo; Lee, Jong-Hyeon; Yi, Kyung-Woo

    2017-09-01

    The waste salt treatment process is required for the reuse of purified salts, and for the disposal of the fission products contained in waste salt during pyroprocessing. As an alternative to existing fission product separation methods, the horizontal zone refining process is used in this study for the purification of waste salt. In order to evaluate the purification ability of the process, three-dimensional simulation is conducted, considering heat transfer, melt flow, and mass transfer. Impurity distributions and decontamination factors are calculated as a function of the heater traverse rate, by applying a subroutine and the equilibrium segregation coefficient derived from the effective segregation coefficients. For multipass cases, 1d solutions and the effective segregation coefficient obtained from three-dimensional simulation are used. In the present study, the topic is not dealing with crystal growth, but the numerical technique used is nearly the same since the zone refining technique was just introduced in the treatment of waste salt from nuclear power industry because of its merit of simplicity and refining ability. So this study can show a new application of single crystal growth techniques to other fields, by taking advantage of the zone refining multipass possibility. The final goal is to achieve the same high degree of decontamination in the waste salt as in zone freezing (or reverse Bridgman) method.

  18. Studies of the suitability of salt domes in east Texas basin for geologic isolation of nuclear wastes

    International Nuclear Information System (INIS)

    Kreitler, C.W.

    1979-01-01

    The suitability of salt domes in the east Texas basin (Tyler basin), Texas, for long-term isolation of nulear wastes is being evaluated. The major issues concern hydrogeologic and tectonic stability of the domes and potential natural resources in the basin. These issues are being approached by integration of dome-specific and regional hydrogeolgic, geologic, geomorphic, and remote-sensing investigations. Hydrogeologic studies are evaluating basinal hydrogeology and ground-water flow around the domes in order to determine the degree to which salt domes may be dissolving, their rates of solution, and the orientation of saline plumes in the fresh-water aquifers. Subsurface geologic studies are being conducted: (1) to determine the size and shape of specific salt domes, the geology of the strata immediately surrounding the domes, and the regional geology of the east Texas basin; (2) to understand the geologic history of dome growth and basin infilling; and (3) to evaluate potential natural resources. Geomorphic and surficial geology studies are determining whether there has been any dome growth or tectonic movement in the basin during the Quaternary. Remote-sensing studies are being conducted to determine: (1) if dome uplift has altered regional lineation patterns in Quaternary sediments; and (2) whether drainage density indicates Quaternary structural movement. On the basis of the screening criteria of Brunton et al (1978), Oakwood and Keechi domes have been chosen as possible candidate domes. Twenty-three domes have been eliminated because of insufficient size, too great a depth to salt, major hydrocarbon production, or previous use (such as liquid propane storage or salt mining or brining). Detailed geologic, hydrogeologic, and geomorphic investigations are now being conducted around Oakwood and Keechi salt domes

  19. Preservation of artifacts in salt mines as a natural analog for the storage of transuranic wastes at the WIPP repository

    International Nuclear Information System (INIS)

    Martell, M.A.; Hansen, F.; Weiner, R.

    1998-01-01

    Use of nature's laboratory for scientific analysis of complex systems is a largely untapped resource for understanding long-term disposal of hazardous materials. The Waste Isolation Pilot Plant (WIPP) in the US is a facility designed and approved for storage of transuranic waste in a salt medium. Isolation from the biosphere must be ensured for 10,000 years. Natural analogs provide a means to interpret the evolution of the underground disposal setting. Investigations of ancient sites where manmade materials have experienced mechanical and chemical processes over millennia provide scientific information unattainable by conventional laboratory methods. This paper presents examples of these pertinent natural analogs, provides examples of features relating to the WIPP application, and identifies potential avenues of future investigations. This paper cites examples of analogical information pertaining to the Hallstatt salt mine in Austria and Wieliczka salt mine in Poland. This paper intends to develop an appreciation for the applicability of natural analogs to the science and engineering of a long-term disposal facility in geomedia

  20. Recovery of soluble chloride salts from the wastewater generated during the washing process of municipal solid wastes incineration fly ash.

    Science.gov (United States)

    Tang, Hailong; Erzat, Aris; Liu, Yangsheng

    2014-01-01

    Water washing is widely used as the pretreatment method to treat municipal solid waste incineration fly ash, which facilitates the further solidification/stabilization treatment or resource recovery of the fly ash. The wastewater generated during the washing process is a kind of hydrosaline solution, usually containing high concentrations of alkali chlorides and sulphates, which cause serious pollution to environment. However, these salts can be recycled as resources instead of discharge. This paper explored an effective and practical recovery method to separate sodium chloride, potassium chloride, and calcium chloride salts individually from the hydrosaline water. In laboratory experiments, a simulating hydrosaline solution was prepared according to composition of the waste washing water. First, in the three-step evaporation-crystallization process, pure sodium chloride and solid mixture of sodium and potassium chlorides were obtained separately, and the remaining solution contained potassium and calcium chlorides (solution A). And then, the solid mixture was fully dissolved into water (solution B obtained). Finally, ethanol was added into solutions A and B to change the solubility of sodium, potassium, and calcium chlorides within the mixed solvent of water and ethanol. During the ethanol-adding precipitation process, each salt was separated individually, and the purity of the raw production in laboratory experiments reached about 90%. The ethanol can be recycled by distillation and reused as the solvent. Therefore, this technology may bring both environmental and economic benefits.

  1. Results of screening activities in salt states prior to the enactment of the Nationall Waste Policy Act

    International Nuclear Information System (INIS)

    Carbiener, W.A.

    1983-01-01

    The identification of potential sites for a nuclear waste repository through screening procedures in the salt states is a well-established, deliberate process. This screening process has made it possible to carry out detailed studies of many of the most promising potential sites, and general studies of all the sites, in anticipation of the siting guidelines specified in the Nuclear Waste Policy Act. The screening work completed prior to the passage of the Act allowed the Secretary of Energy to identify seven salt sites as potentially acceptable under the provisions of Section 116(a) of the Act. These sites were formally identified by letters from Secretary Hodel to the states of Texas, Utah, Mississippi, and Louisiana on February 2, 1983. The potentially acceptable salt sites were in Deaf Smith and Swisher Counties in Texas; Davis and Lavender Canyons in the Gibson Dome location in Utah; Richton and Cypress Creek Domes in Mississippi; and Vacherie Dome in Louisiana. Further screening will include comparison of each potentially acceptable site against disqualification factors and selection of a preferred site in each of the three geohydrologic settings from those remaining, in accordance with the siting guidelines. These steps will be documented in statutory Environmental Assessments prepared for each site to be nominated for detailed characterization. 9 references

  2. Preservation of artifacts in salt mines as a natural analog for the storage of transuranic wastes at the WIPP repository

    Energy Technology Data Exchange (ETDEWEB)

    Martell, M.A.; Hansen, F.; Weiner, R.

    1998-10-01

    Use of nature`s laboratory for scientific analysis of complex systems is a largely untapped resource for understanding long-term disposal of hazardous materials. The Waste Isolation Pilot Plant (WIPP) in the US is a facility designed and approved for storage of transuranic waste in a salt medium. Isolation from the biosphere must be ensured for 10,000 years. Natural analogs provide a means to interpret the evolution of the underground disposal setting. Investigations of ancient sites where manmade materials have experienced mechanical and chemical processes over millennia provide scientific information unattainable by conventional laboratory methods. This paper presents examples of these pertinent natural analogs, provides examples of features relating to the WIPP application, and identifies potential avenues of future investigations. This paper cites examples of analogical information pertaining to the Hallstatt salt mine in Austria and Wieliczka salt mine in Poland. This paper intends to develop an appreciation for the applicability of natural analogs to the science and engineering of a long-term disposal facility in geomedia.

  3. Disposal alternatives and recommendations for waste salt management for repository excavation in the Palo Duro Basin, Texas

    International Nuclear Information System (INIS)

    1987-01-01

    This report documents an evaluation of five alternatives for the disposal of waste salt that would be generated by the construction of a repository for radioactive waste in underground salt deposits at either of two sites in the Palo Duro Basin, Texas. The alternatives include commercial disposal, offsite deep-well injection, disposal in abandoned mines, ocean disposal, and land surface disposal on or off the site. For each alternative a reference case was rated - positive, neutral, or negative - in terms of environmental and dependability factors developed specifically for Texas sites. The factors constituting the environmental checklist relate to water quality impact, water- and land-use conflicts, ecological compatibility, conformity with air quality standards, and aesthetic impact. Factors on the dependability check-list relate to public acceptance, the adequacy of site characterization, permit and licensing requirements, technological requirements, and operational availability. A comparison of the ratings yielded the following viable alternatives, in order of preference: (1) land surface disposal, specifically disposal on tailings piles associated with abandoned potash mines; (2) disposal in abandoned mines, specifically potash mines; and (3) commercial disposal. Approaches to the further study of these three salt management techniques are recommended

  4. Prediction of temperature increases in a salt repository expected from the storage of spent fuel or high-level waste

    International Nuclear Information System (INIS)

    Llewellyn, G.H.

    1978-04-01

    Comparisons in temperature increases incurred from hypothetical storage of 133 MW of 10-year-old spent fuel (SF) or high-level waste (HLW) in underground salt formations have been made using the HEATING5 computer code. The comparisons are based on far-field homogenized models that cover areas of 65 and 25 sq miles for SF and HLW, respectively, and near-field unit-cell models covering respective areas of 610 ft 2 and 400 ft 2 . Preliminary comparisons based on heat loads of 150 kW/acre and 3.5 kW/canister indicated near-field temperature increases about 20% higher for the storage of the spent fuel than for the high-level waste. In these comparisons, it was also found that the thermal energy deposited in the salt after 500 years is about twice the energy deposited by the high-level waste. The thermal load in a repository containing 10-year-old spent fuel was thus limited to 60 kW/acre to obtain comparable far-field thermal effects as obtained in a repository containing 10-year-old high-level waste loaded at 150 kW/acre. Detailed far-field and unit-cell comparisons of transient temperature increases have been made based on these loadings. Unit-cell comparisons were made between a canister containing high-level waste with an initial heat production rate of 2.1 kW and a canister containing a PWR spent fuel assembly producing 0.55 kW. Using a three-dimensional unit-cell model, a maximum salt temperature increase of 260 0 F was calculated for the high-level waste prior to back-filling (5 years after burial), whereas a maximum temperature increase of 110 0 F was calculated for the spent fuel prior to backfilling (25 years after burial). Comparisons were also made between various configurational models for the high-level waste showing the applicability of each model

  5. Use of zinc and copper (I) salts to reduce sulfur and nitrogen impurities during the pyrolysis of plastic and rubber waste to hydrocarbons

    Science.gov (United States)

    Wingfield, Jr., Robert C.; Braslaw, Jacob; Gealer, Roy L.

    1984-01-01

    An improvement in a process for the pyrolytic conversion of rubber and plastic waste to hydrocarbon products which results in reduced levels of nitrogen and sulfur impurities in these products. The improvement comprises pyrolyzing the waste in the presence of at least about 1 weight percent of salts, based on the weight of the waste, preferably chloride or carbonate salts, of zinc or copper (I). This invention was made under contract with or subcontract thereunder of the Department of Energy Contract #DE-AC02-78-ER10049.

  6. Deletion of Cyclooxygenase-2 in the mouse increases arterial blood pressure with no impairment in renal NO production in response to chronic high salt intake

    DEFF Research Database (Denmark)

    Staehr, Mette; Hansen, Pernille B L; Madsen, Kirsten

    2013-01-01

    Experiments were designed to test the hypothesis that COX-2 activity attenuates the blood pressure increase during high NaCl intake by stimulation of eNOS-mediated NO synthesis in the kidney medulla. COX-2(-/-) (C57BL6) and (+/+) mice were fed a diet with 0.004% (LS) or 4% (HS) NaCl for 18 days. ...... pressure during high salt intake and COX-2 activity is not necessary for increased renal NO formation during elevated NaCl intake....... pressure on salt intake and genotype: COX-2(-/-) exhibited higher blood pressure than COX-2(+/+) both on HS and LS intake. COX-2(+/+) littermates displayed an increase in blood pressure on HS vs. LS (102.3±1.1 mmHg vs. 91.9±0.9 mmHg) day and night. The mice exhibited significant blood pressure increases...... during the awake phase (night) that were larger in COX-2(-/-) on HS diet compared to COX-2(+/+). Water intake, diuresis, Na(+) and osmolyte excretions and NOx and cGMP excretions were significantly and similarly elevated with HS in COX-2(-/-) and COX-2(+/+). In summary, C57BL6 mice exhibit a salt intake...

  7. Analysis of water content in salt deposits: its application to radioactive waste storage

    International Nuclear Information System (INIS)

    Cuevas Muller, C. de la.

    1993-01-01

    The salt deposits as radioactive storage medium are analyzed. This report studies the physical-chemical characteristics of water into salts deposits, its implications for the safety of the repository, and the transport water release mechanism. The last part analyzes the geochemical numerical data of correlation analysis, geostatistics analysis and interpretation of statistical data

  8. The potential for using slags activated with near neutral salts as immobilisation matrices for nuclear wastes containing reactive metals

    Science.gov (United States)

    Bai, Y.; Collier, N. C.; Milestone, N. B.; Yang, C. H.

    2011-06-01

    The UK currently uses composite blends of Portland cement and other inorganic cementitious material such as blastfurnace slag and pulverised fuel ash to encapsulate or immobilise intermediate and low level radioactive wastes. Typically levels up 9:1 blast furnace slag:Portland cement or 4:1 pulverised fuel ash:Portland cement are used. Whilst these systems offer many advantages, their high pH causes corrosion of various metallic intermediate level radioactive wastes. To address this issue, lower pH/weakly alkaline cementitious systems have to be explored. While the blast furnace slag:Portland cement system is referred to as a composite cement system, the underlying reaction is actually an indirect activation of the slag hydration by the calcium hydroxide generated by the cement hydration, and by the alkali ions and gypsum present in the cement. However, the slag also can be activated directly with activators, creating a system known as alkali-activated slag. Whilst these activators used are usually strongly alkaline, weakly alkaline and near neutral salts can also be used. In this paper, the potential for using weakly alkaline and near neutral salts to activate slag in this manner is reviewed and discussed, with particular emphasis placed on the immobilisation of reactive metallic nuclear wastes.

  9. Corrosion testing of selected packaging materials for disposal of high-level waste glass in rock salt formations

    International Nuclear Information System (INIS)

    Smailos, E.; Schwarzkopf, W.; Koester, R.; Fiehn, B.; Halm, G.

    1990-05-01

    In previous corrosion studies performed in salt brines, unalloyed steels, Ti 99.8-Pd and Hastelloy C4 have proved to be the most promising materials for long-term resistant packagings to be used in heat-generating waste (vitrified HLW, spent fuel) disposal in rock-salt formations. To characterise the corrosion behaviour of these materials in more detail, further in-depth laboratory-scale and in-situ corrosion studies have been performed in the present study. Besides the above-mentioned materials, also some in-situ investigations of the iron-base materials Ni-Resist D2 and D4, cast iron and Si-cast iron have been carried out in order to complete the results available to date. (orig.) [de

  10. Norepinephrine-evoked salt-sensitive hypertension requires impaired renal sodium chloride cotransporter activity in Sprague-Dawley rats.

    Science.gov (United States)

    Walsh, Kathryn R; Kuwabara, Jill T; Shim, Joon W; Wainford, Richard D

    2016-01-15

    Recent studies have implicated a role of norepinephrine (NE) in the activation of the sodium chloride cotransporter (NCC) to drive the development of salt-sensitive hypertension. However, the interaction between NE and increased salt intake on blood pressure remains to be fully elucidated. This study examined the impact of a continuous NE infusion on sodium homeostasis and blood pressure in conscious Sprague-Dawley rats challenged with a normal (NS; 0.6% NaCl) or high-salt (HS; 8% NaCl) diet for 14 days. Naïve and saline-infused Sprague-Dawley rats remained normotensive when placed on HS and exhibited dietary sodium-evoked suppression of peak natriuresis to hydrochlorothiazide. NE infusion resulted in the development of hypertension, which was exacerbated by HS, demonstrating the development of the salt sensitivity of blood pressure [MAP (mmHg) NE+NS: 151 ± 3 vs. NE+HS: 172 ± 4; P salt-sensitive animals, increased NE prevented dietary sodium-evoked suppression of peak natriuresis to hydrochlorothiazide, suggesting impaired NCC activity contributes to the development of salt sensitivity [peak natriuresis to hydrochlorothiazide (μeq/min) Naïve+NS: 9.4 ± 0.2 vs. Naïve+HS: 7 ± 0.1; P salt-sensitive component of NE-mediated hypertension, while chronic ANG II type 1 receptor antagonism significantly attenuated NE-evoked hypertension without restoring NCC function. These data demonstrate that increased levels of NE prevent dietary sodium-evoked suppression of the NCC, via an ANG II-independent mechanism, to stimulate the development of salt-sensitive hypertension. Copyright © 2016 the American Physiological Society.

  11. Radioactive waste isolation in salt: Peer review of the Fluor Technology, Inc., report and position paper concerning waste emplacement mode and its effect on repository conceptual design

    International Nuclear Information System (INIS)

    Hambley, D.F.; Russell, J.E.; Whitfield, R.G.

    1987-02-01

    Recommendations for revising the Fluor Technology, Inc., draft position paper entitled Evaluation of Waste Emplacement Mode and the final report entitled Waste Package/Repository Impact Study include: reevaluate the relative rankings for the various emplacement modes; delete the following want objectives: maximize ability to locate the package horizon because sufficient flexibility exists to locate rooms in the relatively clean San Andres Unit 4 Salt and maximize far-field geologic integrity during retrieval because by definition the far field will be unaffected by thermal and stress perturbations caused by remining; give greater emphasis to want objectives regarding cost and use of present technology; delete the following statements from pages 1-1 and 1-2 of the draft position paper: ''No thought or study was given to the impacts of this configuration [vertical emplacement] on repository construction or short and long-term performance of the site'' and ''Subsequent salt repository designs adopted the vertical emplacement configuration as the accepted method without further evaluation.''; delete App. E and lines 8-17 of page 1-4 of the draft position paper because they are inappropriate; adopt a formal decision-analysis procedure for the 17 identified emplacement modes; revise App. F of the impact study to more accurately reflect current technology; consider designing the underground layout to take advantage of stress-relief techniques; consider eliminating reference to fuel assemblies <10 yr ''out-of-reactor''; model the temperature distribution, assuming that the repository is constructed in an infinitely large salt body; state that the results of creep analyses must be considered tentative until they can be validated by in situ measurements; and reevaluate the peak radial stresses on the waste package so that the calculated stress conditions more closely approximate expected in situ conditions

  12. Diagnosis and Management of Combined Central Diabetes Insipidus and Cerebral Salt Wasting Syndrome After Traumatic Brain Injury.

    Science.gov (United States)

    Wu, Xuehai; Zhou, Xiaolan; Gao, Liang; Wu, Xing; Fei, Li; Mao, Ying; Hu, Jin; Zhou, Liangfu

    2016-04-01

    Combined central diabetes insipidus and cerebral salt wasting syndrome after traumatic brain injury (TBI) is rare, is characterized by massive polyuria leading to severe water and electrolyte disturbances, and usually is associated with very high mortality mainly as a result of delayed diagnosis and improper management. We retrospectively reviewed the clinical presentation, management, and outcomes of 11 patients who developed combined central diabetes insipidus and cerebral salt wasting syndrome after traumatic brain injury to define distinctive features for timely diagnosis and proper management. The most typical clinical presentation was massive polyuria (10,000 mL/24 hours or >1000 mL/hour) refractory to vasopressin alone but responsive to vasopressin plus cortisone acetate. Other characteristic presentations included low central venous pressure, high brain natriuretic peptide precursor level without cardiac dysfunction, high 24-hour urine sodium excretion and hypovolemia, and much higher urine than serum osmolarity; normal serum sodium level and urine specific gravity can also be present. Timely and adequate infusion of sodium chloride was key in treatment. Of 11 patients, 5 had a good prognosis 3 months later (Extended Glasgow Outcome Scale score ≥6), 1 had an Extended Glasgow Outcome Scale score of 4, 2 died in the hospital of brain hernia, and 3 developed a vegetative state. For combined diabetes insipidus and cerebral salt wasting syndrome after traumatic brain injury, massive polyuria is a major typical presentation, and intensive monitoring of fluid and sodium status is key for timely diagnosis. To achieve a favorable outcome, proper sodium chloride supplementation and cortisone acetate and vasopressin coadministration are key. Copyright © 2016 Elsevier Inc. All rights reserved.

  13. Thermomechanical effects of the salt rock on the solidified waste product during ultimate stoage of radioactive waste

    International Nuclear Information System (INIS)

    Schoen, R.

    1981-01-01

    The thermal stresses in the salt to be expected in the elastic case are very much reduced by the viscous behavior of the salt rock. The occurrence of tensile stresses may be prevented by reducing the differential temperatures by means of a decrease of the mould heat rate and/or the mechanical behavior of the glass as well as design measures. As far as the mechanical aspect is concerned thicker coverings have no positive effect on the stress in the glass. In the course of time the three principal stresses in the salt rock are matching. At the terminal point of the reference calculations these stresses amount to 12.5 MPa and 15 MPa in the horizontal and vertical direction respectively. (DG) [de

  14. Radiation damage studies on natural rock salt from various geological localities of interest to the radioactive waste disposal program

    International Nuclear Information System (INIS)

    Levy, P.W.

    1981-01-01

    As part of a program to investigate radiation damage in geological materials of interest to the radioactive waste disposal program, radiation damage, particularly radiation induced sodium metal colloid formation, has been studied in 14 natural rock salt samples. All measurements were made with equipment for making optical absorption and other measurements on samples, in a temperature controlled irradiation chamber, during and after 0.5 to 3.0 MeV electron irradiation. Samples were chosen for practical and scientific purposes, from localities that are potential repository sites and from different horizons at certain localities

  15. Preliminary Technical and Legal Evaluation of Disposing of Nonhazardous Oil Field Waste into Salt Caverns

    National Research Council Canada - National Science Library

    Veil, John

    1996-01-01

    .... These caverns are either created incidentally as a result of salt recovery or intentionally to create an underground chamber that can be used for storing hydrocarbon products or compressed air...

  16. The HAW-project: Demonstration facility for the disposal of high-level waste in salt

    International Nuclear Information System (INIS)

    Rothfuchs, T.; Duijves, K.A.; Mueller-Lyda, I.

    1990-04-01

    To satisfy the test objectives thirty highly radioactive canisters containing the radionuclides Cs-137 and Sr-90 will be emplaced in six boreholes located in two test galleries at the 800 m-level in the Asse salt mine. For handling of the radioactive canisters and their emplacement into the boreholes a system consisting of a transport cask, a transport vehicle, a disposal machine, and of a borehole slider has been developed. The actual scientific investigation programme is based on the estimation and observation of the interaction between the radioactive canisters and the rock salt. This programme includes measurement of thermally and radiolytically induced water and gas release from the rock salt and the radiolytical decomposition of salt minerals. Also the thermally induced stress and deformation fields in the surrounding rock mass will be investigated carefully. (orig./DG)

  17. Regulatory Framework for Salt Waste Disposal and Tank Closure at the Savannah River Site - 13663

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, Steve; Dickert, Ginger [Savannah River Remediation LLC, Savannah River Site, Aiken, SC 29808 (United States)

    2013-07-01

    The end of the Cold War has left a legacy of approximately 37 million gallons of radioactive waste in the aging waste tanks at the Department of Energy's Savannah River Site (SRS). A robust program is in place to remove waste from these tanks, treat the waste to separate into a relatively small volume of high-level waste and a large volume of low-level waste, and to actively dispose of the low-level waste on-site and close the waste tanks and associated ancillary structures. To support performance-based, risk-informed decision making and to ensure compliance with all regulatory requirements, the U.S. Department of Energy (DOE) and its current and past contractors have worked closely with the South Carolina Department of Health and Environmental Control (SCDHEC), the U.S. Environmental Protection Agency (EPA) and the Nuclear Regulatory Commission (NRC) to develop and implement a framework for on-site low-level waste disposal and closure of the SRS waste tanks. The Atomic Energy Act of 1954, as amended, provides DOE the authority to manage defense-related radioactive waste. DOE Order 435.1 and its associated manual and guidance documents detail this radioactive waste management process. The DOE also has a requirement to consult with the NRC in determining that waste that formerly was classified as high-level waste can be safely managed as either low-level waste or transuranic waste. Once DOE makes a determination, NRC then has a responsibility to monitor DOE's actions in coordination with SCDHEC to ensure compliance with the Title 10 Code of Federal Regulations Part 61 (10CFR61), Subpart C performance objectives. The management of hazardous waste substances or components at SRS is regulated by SCDHEC and the EPA. The foundation for the interactions between DOE, SCDHEC and EPA is the SRS Federal Facility Agreement (FFA). Managing this array of requirements and successfully interacting with regulators, consultants and stakeholders is a challenging task but

  18. Microbial Influence on the Performance of Subsurface, Salt-Based Radioactive Waste Repositories. An Evaluation Based on Microbial Ecology, Bioenergetics and Projected Repository Conditions

    International Nuclear Information System (INIS)

    Swanson, J.S.; Reed, D.T.; Cherkouk, A.; Arnold, T.; Meleshyn, A.; Patterson, Russ

    2018-01-01

    For the past several decades, the Nuclear Energy Agency Salt Club has been supporting and overseeing the characterisation of rock salt as a potential host rock for deep geological repositories. This extensive evaluation of deep geological settings is aimed at determining - through a multidisciplinary approach - whether specific sites are suitable for radioactive waste disposal. Studying the microbiology of granite, basalt, tuff, and clay formations in both Europe and the United States has been an important part of this investigation, and much has been learnt about the potential influence of microorganisms on repository performance, as well as about deep subsurface microbiology in general. Some uncertainty remains, however, around the effects of microorganisms on salt-based repository performance. Using available information on the microbial ecology of hyper-saline environments, the bioenergetics of survival under high ionic strength conditions and studies related to repository microbiology, this report summarises the potential role of microorganisms in salt-based radioactive waste repositories

  19. Performance assessment of confinements for medium-level and α-contaminated waste. PACOMA project. Rock salt option

    International Nuclear Information System (INIS)

    Hirsekorn, R.P; Nies, A.; Rausch, H.; Storck, R.

    1991-03-01

    The objective of the contribution to the PACOMA project is to develop and demonstrate procedures for radiological safety of repositories in salt domes. An analogue study is performed by the Netherlands Energy Research Foundation ECN, where alternative disposal concepts in different salt formations were investigated. It is discussed, how far appropriate choice of the repository design parameters can improve the whole systems. The research covers deterministic calculations for three scenarios, the normal evolution scenario with subrosion of the salt dome, the combined brine intrusion scenario with brine intrusion from brine pockets and via an anhydrite vein, and the human intrusion scenario of solution mining of a storage cavern. For the combined brine intrusion scenario alternative waste inventories, different disposal concepts, variants of the layout of dams and sealings are investigated, and results obtained from variations of parameter values are discussed. Additionally, comprehensive probabilistic calculations have been carried out with the help of a Monte-Carlo simulation. Results are discussed in form of an uncertainty analysis of the maximum dose and global sensitivity studies of system parameters. The assessments main result is, that the reference case, where the reference repository design and the reference disposal concept are applied, deterministic calculations with best estimate values as well as probabilistic calculations do not manifest unacceptable risk. Investigation of alternative concepts and design variants indicate a high potential for system optimization. (orig./HP)

  20. Biosphere transport and radiation dose calculations resulting from radioactive waste stored in deep salt formation (PACOMA-project)

    International Nuclear Information System (INIS)

    Jong, E.J. de; Koester, H.W.; Vries, W.J. de; Lembrechts, J.F.

    1990-03-01

    Parts are presented of the results of a safety-assessment study of disposal of medium and low level radioactive waste in salt formations in the Netherlands. The study concerns several disposal concepts for 2 kinds of salt formation, a deep dome and a shallow dome. 7 cases were studied with the same Dutch inventory and 1 with a reference inventory R, in order to compare results with those of other PACOMA participants. The total activity of the reference inventory R is 30 percent lower than the Dutch inventory, but some long living nuclides such as I-129, Np-237 and U-238 have a considerably higher activity. This reference inventor R has been combined with the disposal concept of mined cavities in a shallow salt dome. In each case. the released fraction of stored radio-nuclides moves gradually with water through the geosphere to the bio-sphere where it enters a river. River water is used for sprinkler irrigation and for drinking by man and livestock. The dispersal of the radionuclides into the biosphere is calculated with the BIOS program of the NRPB. Subroutines linked to the program add doses via different pathways to obtain a maximum individual dose, a collective dose and an integrated collective dose. This study presents results of these calculations. (author). 11 refs.; 39 figs.; 111 tabs

  1. Geohydrology of the northern Louisiana salt-dome basin pertinent to the storage of radioactive wastes; a progress report

    Science.gov (United States)

    Hosman, R.L.

    1978-01-01

    Salt domes in northern Louisiana are being considered as possible storage sites for nuclear wastes. The domes are in an area that received regional sedimentation through early Tertiary (Eocene) time with lesser amounts of Quaternary deposits. The Cretaceous-Tertiary accumulation is a few thousand feet thick; the major sands are regional aquifers that extend far beyond the boundaries of the salt-dome basin. Because of multiple aquifers, structural deformation, and variations in the hydraulic characteristics of cap rock, the ground-water hydrology around a salt dome may be highly complex. The Sparta Sand is the most productive and heavily used regional aquifer. It is either penetrated by or overlies most of the domes. A fluid entering the Sparta flow system would move toward one of the pumping centers, all at or near municipalities that pump from the Sparta. Movement could be toward surface drainage where local geologic and hydrologic conditions permit leakage to the surface or to a surficial aquifer. (Woodard-USGS)

  2. National waste terminal storage repository in a bedded salt formation for spent unreprocessed fuel. Volume I. Conceptual design report

    International Nuclear Information System (INIS)

    1978-12-01

    In February 1976, the Energy Research and Development Administration (ERDA), now the Department of Energy (DOE), established a National Waste Terminal Storage (NWTS) program. As a part of this program, two parallel conceptual design efforts were initiated in January 1977. One was for deep geologic storage, in domed salt, of high level waste resulting from the reprocessing of spent fuel. The other was for deep geologic storage of unreprocessed spent fuel in bedded salt. These two concepts are identified as NWTS Repository 1 and Repository 2, respectively. Repository 2 (NWTSR2) is the concept which is covered by this Conceptual Design Report. Volume I of the conceptual design report contains the following information: physical description of the report; project purpose and justification; principal safety, fire, and health hazards; environmental impact considerations; quality assurance considerations; assessment of operational interfaces; assessment of research and development interfaces; project schedule; proposed method of accomplishment; summary cost estimate; and outline specifications. The conceptual design for Repository 2 was developed in sufficient detail to permit determination of scope, engineering feasibility, schedule, and cost estimates, all of which are necessary for planning and budgeting the project

  3. Salton Sea Geothermal Field, California, as a near-field natural analog of a radioactive waste repository in salt

    International Nuclear Information System (INIS)

    Elders, W.A.; Cohen, L.H.

    1983-11-01

    Since high concentrations of radionuclides and high temperatures are not normally encountered in salt domes or beds, finding an exact geologic analog of expected near-field conditions in a mined nuclear waste repository in salt will be difficult. The Salton Sea Geothermal Field, however, provides an opportunity to investigate the migration and retardation of naturally occurring U, Th, Ra, Cs, Sr and other elements in hot brines which have been moving through clay-rich sedimentary rocks for up to 100,000 years. The more than thirty deep wells drilled in this field to produce steam for electrical generation penetrate sedimentary rocks containing concentrated brines where temperatures reach 365 0 C at only 2 km depth. The brines are primarily Na, K, Ca chlorides with up to 25% of total dissolved solids; they also contain high concentrations of metals such as Fe, Mn, Li, Zn, and Pb. This report describes the geology, geophysics and geochemistry of this system as a prelude to a study of the mobility of naturally occurring radionuclides and radionuclide analogs within it. The aim of this study is to provide data to assist in validating quantitative models of repository behavior and to use in designing and evaluating waste packages and engineered barriers. 128 references, 33 figures, 13 tables

  4. Salton Sea Geothermal Field, California, as a near-field natural analog of a radioactive waste repository in salt

    Energy Technology Data Exchange (ETDEWEB)

    Elders, W.A.; Cohen, L.H.

    1983-11-01

    Since high concentrations of radionuclides and high temperatures are not normally encountered in salt domes or beds, finding an exact geologic analog of expected near-field conditions in a mined nuclear waste repository in salt will be difficult. The Salton Sea Geothermal Field, however, provides an opportunity to investigate the migration and retardation of naturally occurring U, Th, Ra, Cs, Sr and other elements in hot brines which have been moving through clay-rich sedimentary rocks for up to 100,000 years. The more than thirty deep wells drilled in this field to produce steam for electrical generation penetrate sedimentary rocks containing concentrated brines where temperatures reach 365/sup 0/C at only 2 km depth. The brines are primarily Na, K, Ca chlorides with up to 25% of total dissolved solids; they also contain high concentrations of metals such as Fe, Mn, Li, Zn, and Pb. This report describes the geology, geophysics and geochemistry of this system as a prelude to a study of the mobility of naturally occurring radionuclides and radionuclide analogs within it. The aim of this study is to provide data to assist in validating quantitative models of repository behavior and to use in designing and evaluating waste packages and engineered barriers. 128 references, 33 figures, 13 tables.

  5. Thermal Properties of LiCl-KCl Molten Salt for Nuclear Waste Separation

    International Nuclear Information System (INIS)

    Sridharan, Kumar; Allen, Todd; Anderson, Mark; Simpson, Mike

    2012-01-01

    This project addresses both practical and fundamental scientific issues of direct relevance to operational challenges of the molten LiCl-KCl salt pyrochemical process, while providing avenues for improvements in the process. In order to understand the effects of the continually changing composition of the molten salt bath during the process, the project team will systematically vary the concentrations of rare earth surrogate elements, lanthanum, cerium, praseodymium, and neodymium, which will be added to the molten LiCl-KCl salt. They will also perform a limited number of focused experiments by the dissolution of depleted uranium. All experiments will be performed at 500 deg C. The project consists of the following tasks. Researchers will measure density of the molten salts using an instrument specifically designed for this purpose, and will determine the melting points with a differential scanning calorimeter. Knowledge of these properties is essential for salt mass accounting and taking the necessary steps to prevent melt freezing. The team will use cyclic voltammetry studies to determine redox potentials of the rare earth cations, as well as their diffusion coefficients and activities in the molten LiCl-KCl salt. In addition, the team will perform anodic stripping voltammetry to determine the concentration of the rare earth elements and their solubilities, and to develop the scientific basis for an on-line diagnostic system for in situ monitoring of the cation species concentration (rare earths in this case). Solubility and activity of the cation species are critically important for the prediction of the salt's useful lifetime and disposal

  6. Thermal Properties of LiCl-KCl Molten Salt for Nuclear Waste Separation

    Energy Technology Data Exchange (ETDEWEB)

    Sridharan, Kumar [Univ. of Wisconsin, Madison, WI (United States); Allen, Todd [Univ. of Wisconsin, Madison, WI (United States); Anderson, Mark [Univ. of Wisconsin, Madison, WI (United States); Simpson, Mike [Idaho National Lab., (United States)

    2012-11-30

    This project addresses both practical and fundamental scientific issues of direct relevance to operational challenges of the molten LiCl-KCl salt pyrochemical process, while providing avenues for improvements in the process. In order to understand the effects of the continually changing composition of the molten salt bath during the process, the project team will systematically vary the concentrations of rare earth surrogate elements, lanthanum, cerium, praseodymium, and neodymium, which will be added to the molten LiCl-KCl salt. They will also perform a limited number of focused experiments by the dissolution of depleted uranium. All experiments will be performed at 500 deg C. The project consists of the following tasks. Researchers will measure density of the molten salts using an instrument specifically designed for this purpose, and will determine the melting points with a differential scanning calorimeter. Knowledge of these properties is essential for salt mass accounting and taking the necessary steps to prevent melt freezing. The team will use cyclic voltammetry studies to determine redox potentials of the rare earth cations, as well as their diffusion coefficients and activities in the molten LiCl-KCl salt. In addition, the team will perform anodic stripping voltammetry to determine the concentration of the rare earth elements and their solubilities, and to develop the scientific basis for an on-line diagnostic system for in situ monitoring of the cation species concentration (rare earths in this case). Solubility and activity of the cation species are critically important for the prediction of the salt's useful lifetime and disposal.

  7. Radiolytic bubble formation and level changes in simulated high-level waste salts and sludges -- application to Savannah River Site and Hanford Storage tanks

    International Nuclear Information System (INIS)

    Walker, D.D.; Crawford, C.L.; Bibler, N.E.

    1993-01-01

    Radiolytically-produced bubbles of trapped gas are observed in simulated high-level waste (HLW) damp salt cake exposed to Co-60 gamma radiation. As the damp salt cake is irradiated, its volume increases due to the formation of trapped gas bubbles. Based on the increase in volume, the rate of trapped gas generation varies between 0.04 and 0.2 molecules/100 eV of energy deposited in the damp salt cake. The maximum volume of trapped gas observed in experiments is in the range 21--26 vol %. After reaching these volumes, the gas bubbles begin to escape. The generated gas includes hydrogen, oxygen, and nitrous oxide. The ratio in which these components are produced depends on the composition of the waste. Nitrous oxide production increases with the amount of sodium nitrite. Gases trapped by this mechanism may account for some of the observed level changes in Savannah River Site and Hanford waste tanks

  8. Accumulated energy determination in salts rocks irradiated by means of thermoluminescence techniques: application to the high level radioactive wastes repositories analysis

    International Nuclear Information System (INIS)

    Dies, J.; Ortega. J.; Tarrasa. F.; Cuevas, C.

    1995-01-01

    The report summarizes the study carried out to develop the radiation effects on salt rocks in order to repository the high level radioactive wastes. The study is structured into 3 main aspects: 1.- Analysis of irradiation experiences in Haw project of Pet ten reactor. 2.- Irradiation of salt sample of CESAR industrial irradiator. 3.- Correlation study between the accumulated energy, termoluminescence answer and the defect concentration

  9. Alteration of rhyolitic (volcanic) glasses in natural Bolivian salt lakes. - Natural analogue for the behavior of radioactive waste glasses in rock salt repositories

    International Nuclear Information System (INIS)

    Abdelouas, A.

    1996-06-01

    Alteration experiments with the R7T7 glass in three salt brines, saturated respectively in MgCl 2 , MgCl 2 -CaCl 2 and NaCl, showed that the solubilities of most radionuclides are controlled by the secondary phases. Nd, La, and Pr are trapped in powellite, Ce in cerianite, U in coffinite, and Sr is partially immobilized in barite. There is a good similarity between the secondary phases formed experimentally on volcanic glasses and the R7T7 glass altered in MgCl 2 CaCl 2 -saturated brine (formation of hydrotalcite and chlorite-serpentine at short-term and saponite at long-term). These results support the use of volcanic glasses alteration patterns in Mg-rich solutions (seawater, brines) to understand the long-term behavior of nuclear waste glasses and to evaluate the stability of the secondary phases. The study of the sediments of Uyuni (Bolivia) showed that the corrosion rate of the rhyolitic glass in brines at 10 C is 12 to 30 time lower than those of rhyolitic glasses altered in high dilute conditions. The neoformed phases in the sediments are: Smectite, alunite, pyrite, barite, celestite and cerianite. The low alteration rate of rhyolitic glasses in brines and the formation of secondary phases such as smectite, barite and cerianite (also formed during the experimental alteration of the R7T7 glass), permit us to expect the low alteration of nuclear waste glasses at long-term in brines and the trapping of certain radionuclides in secondary phases. (orig.) [de

  10. Testing of Air Pulse Agitators to Support Design of Savannah River Site Highly Radioactive Processing at the Salt Waste Processing Facility

    International Nuclear Information System (INIS)

    Gallego, R.M.; Stephens, A.B.; Wilkinson, R.H.; Dev, H.; Suggs, P.C.

    2006-01-01

    The Salt Waste Processing Facility (SWPF) is intended to concentrate the highly radioactive constituents from waste salt solutions at the Savannah River Site (SRS). Air Pulse Agitators (APAs) were selected for process mixing in high-radiation locations at the SWPF. This technology has the advantage of no moving parts within the hot cell, eliminating potential failure modes and the need for maintenance within the high-radiation environment. This paper describes the results of APA tests performed to gain operational and performance data for the SWPF design. (authors)

  11. Radioactive waste isolation in salt: peer review of the Office of Nuclear Waste Isolation's reports on preferred repository sites within the Palo Duro Basin, Texas

    International Nuclear Information System (INIS)

    Fenster, D.; Edgar, D.; Gonzales, S.; Domenico, P.; Harrison, W.; Engelder, T.; Tisue, M.

    1984-04-01

    Documents are being submitted to the Salt Repository Project Office (SRPO) of the US Department of Energy (DOE) by Battelle Memorial Institute's Office of Nuclear Waste Isolation (ONWI) to satisfy milestones of the Salt Repository Project of the Civilian Radioactive Waste Management Program. Some of these documents are being reviewed by multidisciplinary groups of peers to ensure DOE of their adequacy and credibility. Adequacy of documents refers to their ability to meet the standards of the US Nuclear Regulatory Commission, as enunciated in 10 CFR 60, and the requirements of the National Environmental Policy Act and the Nuclear Waste Policy Act of 1982. Credibility of documents refers to the validity of the assumptions, methods, and conclusions, as well as to the completeness of coverage. This report summarizes Argonne's review of ONWI's two-volume draft report entitled Identification of Preferred Sites within the Palo Duro Basin: Vol. 1 - Palo Duro Location A, and Vol. 2 - Palo Duro Location B, dated January 1984. Argonne was requested by DOE to review these documents on January 17 and 24, 1984 (see App. A). The review procedure involved obtaining written comments on the reports from three members of Argonne's core peer review staff and three extramural experts in related research areas. The peer review panel met at Argonne on February 6, 1984, and reviewer comments were integrated into this report by the review session chairman, with the assistance of Argonne's core peer review staff. All of the peer review panelists concurred in the way in which their comments were represented in this report (see App. B). A letter report and a draft of this report were sent to SRPO on February 10, 1984, and April 17, 1984, respectively. 5 references

  12. The Herfa-Neurode hazardous waste repository in bedded salt as an operating model for safe mixed waste disposal

    International Nuclear Information System (INIS)

    Rempe, N.T.

    1991-01-01

    For 18 years, The Herfa-Neurode underground repository has demonstrated the environmentally sound disposal of hazardous waste in a former potash mine. Its principal characteristics make it an excellent analogue to the Waste Isolation Pilot Plant (WIPP). The Environmental Protection Agency has ruled in its first conditional no-migration determination that is reasonably certain that no hazardous constituents of the mixed waste, destined for the WIPP during its test phase, will migrate from the site for up to ten years. Knowledge of and reference to the Herfa-Neurode operating model may substantially improve the no-migration variance petition for the WIPP's disposal phase and thereby expedite its approval. 2 refs., 1 fig., 1 tab

  13. A Glass-Ceramic Waste Forms for the Immobilization of Rare Earth Oxides from the Pyroprocessing Waste salt

    International Nuclear Information System (INIS)

    Ahn, Byung-Gil; Park, Hwan-Seo; Kim, Hwan-Young; Kim, In-Tae

    2008-01-01

    The fission product of rare earth (RE) oxide wastes are generates during the pyroprocess . Borosilicate glass or some ceramic materials such as monazite, apatite or sodium zirconium phosphate (NZP) have been a prospective host matrix through lots of experimental results. Silicate glasses have long been the preferred waste form for the immobilization of HLW. In immobilization of the RE oxides, the developed process on an industrial scale involves their incorporation into a glass matrix, by melting under 1200 ∼ 1300 .deg. C. Instead of the melting process, glass powder sintering is lower temperature (∼ 900 .deg. C) required for the process which implies less demanding conditions for the equipment and a less evaporation of volatile radionuclides. This study reports the behaviors, direct vitrification of RE oxides with glass frit, glass powder sintering of REceramic with glass frit, formation of RE-apatite (or REmonazite) ceramic according to reaction temperature, and the leach resistance of the solidified waste forms

  14. Tank Waste Transport Stability: Summary of Slurry and Salt-Solution Studies for FY 2001

    Energy Technology Data Exchange (ETDEWEB)

    Welch, T.D.

    2002-06-07

    Despite over 50 years of experience in transporting radioactive tank wastes to and from equipment and tanks at the Department of Energy's Hanford, Savannah River, and Oak Ridge sites, waste slurry transfer pipelines and process piping become plugged on occasion. At Hanford, several tank farm pipelines are no longer in service because of plugs. At Savannah River, solid deposits in the outlet line of the 2H evaporator have resulted in an unplanned extended downtime. Although waste transfer criteria and guidelines intended to prevent pipeline plugging are in place, they are not always adequate. To avoid pipeline plugging in the future, other factors that are not currently embodied in the transfer criteria may need to be considered. The work summarized here is being conducted to develop a better understanding of the chemical and waste flow dynamics during waste transfer. The goal is to eliminate pipeline plugs by improving analysis and engineering tools in the field that incorporate this understanding.

  15. The waste isolation pilot plant. Permanent isolation of defense transuranic waste in deep geologic salt. A national solution and international model

    International Nuclear Information System (INIS)

    Franco, Jose; Van Luik, Abraham

    2015-01-01

    The Waste Isolation Pilot Plant is located about 42 kilometers from the city of Carlsbad, New Mexico. It is an operating deep geologic repository in bedded salt 657 meters below the surface of the Chihuahuan desert. Since its opening in March of 1999, it has received about 12,000 shipments totaling about 91,000 cubic meters of defense related transuranic (TRU) wastes. Twenty-two sites have been cleaned up of their defense-legacy TRU waste. The WIPP's shipping program has an untarnished safety record and its trucks and trailers have safely traveled the equivalent of about 60 round-trips to the Moon. WIPP received, and deserved, a variety of safety accolades over its nearly 15 year working life. In February of 2014, however, two incidents resulted in a major operational suspension and reevaluation of its safety systems, processes and equipment. The first incident was an underground mining truck fire, followed nine days later by an airborne radiation release incident. Accident Investigation Board (AIB) reports on both incidents point to failures of plans, procedures and persons. The AIB recommendations for recovery from both these incidents are numerous and are being carefully implemented. One major recommendation is to no longer have different maintenance and safety requirements for nuclear handling equipment and mining equipment. Maintenance and cleanliness of mining equipment was cited as a contributing cause to the underground fire, and the idea that there can be lesser rigor in taking care of mining equipment, when it is being operated in the same underground space as the waste handling equipment, is not tenable. At some point in the future, the changes made in response to these two incidents will be seen as a valuable lesson learned on behalf of future repository programs. WIPP will once again be seen as a ''pilot'' in the nautical sense, in terms of 'showing the way' - the way to a national and international radioactive waste

  16. The waste isolation pilot plant. Permanent isolation of defense transuranic waste in deep geologic salt. A national solution and international model

    Energy Technology Data Exchange (ETDEWEB)

    Franco, Jose; Van Luik, Abraham [US Department of Energy, Carlsbad, NM (United States). Carlsbad Field Office

    2015-07-01

    The Waste Isolation Pilot Plant is located about 42 kilometers from the city of Carlsbad, New Mexico. It is an operating deep geologic repository in bedded salt 657 meters below the surface of the Chihuahuan desert. Since its opening in March of 1999, it has received about 12,000 shipments totaling about 91,000 cubic meters of defense related transuranic (TRU) wastes. Twenty-two sites have been cleaned up of their defense-legacy TRU waste. The WIPP's shipping program has an untarnished safety record and its trucks and trailers have safely traveled the equivalent of about 60 round-trips to the Moon. WIPP received, and deserved, a variety of safety accolades over its nearly 15 year working life. In February of 2014, however, two incidents resulted in a major operational suspension and reevaluation of its safety systems, processes and equipment. The first incident was an underground mining truck fire, followed nine days later by an airborne radiation release incident. Accident Investigation Board (AIB) reports on both incidents point to failures of plans, procedures and persons. The AIB recommendations for recovery from both these incidents are numerous and are being carefully implemented. One major recommendation is to no longer have different maintenance and safety requirements for nuclear handling equipment and mining equipment. Maintenance and cleanliness of mining equipment was cited as a contributing cause to the underground fire, and the idea that there can be lesser rigor in taking care of mining equipment, when it is being operated in the same underground space as the waste handling equipment, is not tenable. At some point in the future, the changes made in response to these two incidents will be seen as a valuable lesson learned on behalf of future repository programs. WIPP will once again be seen as a ''pilot'' in the nautical sense, in terms of 'showing the way' - the way to a national and international radioactive waste

  17. A analysis of molten salt separation system for nuclear wastes transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, In Soon; Park, Byung Gi [Seoul National University, Seoul (Korea, Republic of); Kim, Kwang Bum; Kwon, Ou Sung [Yonsei University, Seoul (Korea, Republic of)

    1997-07-01

    Typical molten salt separation is ANL-IFR pyroprocessing and ORNL-MSRE pyroprocessing. IFR pyroprocessing is based on Chloride chemistry and electrorefining. MSRE pyroprocessing is base on fluoride chemistry and reductive extraction. Major technologies of molten salt separation are electrorefining, electrowining, reductive extraction, and oxide reduction. Common characteristics of this technologies is to utilize reduction-oxidation phenomena in molten salt. Electrorefining process is modeled on the basis of diffusion layer theory and Butler-Volmor relation. This model is numerically solved by LSODA package. To acquire the technology of electrorefining process, 3-electrode electrochemical cell is developed where electrolyte is 500 degree C LiCl-KCl eutectic molten salt, working electrodes are Ni and Au, and reference electrode is Ag/AgCl. We have investigated the stable potential range using cyclic voltammogram of Ni electrode. We have measured steady state polarization curve of Ni electrode. Then corrosion potential of Ni electrode is -0.38V{sub Ag/AgCl} and corrosion current is 1.23 x 10{sup -4} A/cm{sup 2}. 12 refs., 6 tabs., 24 figs. (author)

  18. Comments on a letter by George D. DeBuchananne (US Geological Survey) regarding the use of salt domes for high-level waste disposal

    International Nuclear Information System (INIS)

    1984-08-01

    The US Geological Survey (USGS) concluded in a letter to the US Department of Energy, dated March 7, 1981, that subsurface geologic conditions in bedded salt are more predictable and less complex than those in domal salt. This predictability is equated with the relative suitability of bedded and domal salt as repository host media. This report comments on the USGS letter. The key points made are as follows: Complexities which may exist in the geologic setting of a salt dome (or other potential host medium) should not a priori preclude the dome from being an acceptable host medium for a high-level waste (HLW) repository. Predictability, as used by the USGS, focused on the spatial extrapolation of information on geologic conditions and should not be confused with predicting the performance of a repository. Notwithstanding the general characteristics of bedded and domal salt, there are salt domes whose individual characteristics should make them as acceptable as potential bedded salt areas for HLW repository sites. Complexities which may occur in the geologic setting of a salt dome can be explored and characterized with sufficient accuracy by available techniques

  19. Technetium in alkaline, high-salt, radioactive tank waste supernate: Preliminary characterization and removal

    International Nuclear Information System (INIS)

    Blanchard, D.L. Jr.; Brown, G.N.; Conradson, S.D.

    1997-01-01

    This report describes the initial work conducted at Pacific Northwest National Laboratory to study technetium (Tc) removal from Hanford tank waste supernates and Tc oxidation state in the supernates. Filtered supernate samples from four tanks were studied: a composite double shell slurry feed (DSSF) consisting of 70% from Tank AW-101, 20% from AP-106, and 10% from AP-102; and three complexant concentrate (CC) wastes (Tanks AN-107, SY-101, ANS SY-103) that are distinguished by having a high concentration of organic complexants. The work included batch contacts of these waste samples with Reillex trademark-HPQ (anion exchanger from Reilly Industries) and ABEC 5000 (a sorbent from Eichrom Industries), materials designed to effectively remove Tc as pertechnetate from tank wastes. A short study of Tc analysis methods was completed. A preliminary identification of the oxidation state of non-pertechnetate species in the supernates was made by analyzing the technetium x-ray absorption spectra of four CC waste samples. Molybdenum (Mo) and rhenium (Re) spiked test solutions and simulants were tested with electrospray ionization-mass spectrometry to evaluate the feasibility of the technique for identifying Tc species in waste samples

  20. Incineration: why this may be the most environmentally sound method of renal healthcare waste disposal.

    Science.gov (United States)

    James, Ray

    2010-09-01

    The environment and 'green' issues are currently being promoted in the healthcare sector through recently launched initiatives. This paper considers aspects of healthcare waste management, with particular reference to waste generated in dialysis units. With dialysis being dependent upon large amounts of disposables, it generates considerable volumes of waste. This paper focuses upon a typical haemodialysis unit, evaluating and quantifying the volumes and categories of waste generated. Each haemodialysis patient on thrice weekly dialysis generates some 323 kg per year of waste, of which 271 kg is classified as clinical. This equates to 1626 kg of (solid) clinical waste per dialysis bed, which is around three times the volume of clinical waste generated per general hospital bed. Waste disposal routes are considered and this suggests that present healthcare waste paradigms are outmoded. They do not allow for flexible approaches to solving what is a dynamic problem, and there is a need for new thinking models in terms of managing the unsustainable situation of disposal in constantly growing landfills. Healthcare waste management must be considered not only in terms of the environmental impact and potential long-term health effects, but also in terms of society's future energy requirements.

  1. Cyclooxygenase 2 and neuronal nitric oxide synthase expression in the renal cortex are not interdependent in states of salt deficiency

    DEFF Research Database (Denmark)

    Castrop, H; Kammerl, M; Mann, Birgitte

    2000-01-01

    Neuronal nitric oxide synthase (nNOS) and cyclooxygenase-2 (COX-2) expression in the kidney are localized to the cortical thick ascending limb of the loop of Henle (cTALH), including the macula region, and increase after salt restriction. Because of the similar localization and regulation of n...... excretion. These findings suggest that under these conditions the control of nNOS and COX-2 gene expression in the macula densa regions of the kidney cortex are not dependent on each other....

  2. Review of information on the radiation chemistry of materials around waste canisters in salt and assessment of the need for additional experimental information

    Energy Technology Data Exchange (ETDEWEB)

    Jenks, G.H.; Baes, C.F. Jr.

    1980-03-01

    The brines, vapors, and salts precipitated from the brines will be exposed to gamma rays and to elevated temperatures in the regions close to a waste package in the salt. Accordingly, they will be subject to changes in composition brought about by reactions induced by the radiations and heat. This report reviews the status of information on the radiation chemistry of brines, gases, and solids which might be present around a waste package in salt and to assess the need for additional laboratory investigations on the radiation chemistry of these materials. The basic aspects of the radiation chemistry of water and aqueous solutions, including concentrated salt solutions, were reviewed briefly and found to be substantially unchanged from those presented in Jenks's 1972 review of radiolysis and hydrolysis in salt-mine brines. Some additional information pertaining to the radiolytic yields and reactions in brine solutions has become available since the previous review, and this information will be useful in the eventual, complete elucidation of the radiation chemistry of the salt-mine brines. 53 references.

  3. Geology, hydrology, thickness and quality of salt at three alternate sites for disposal of radioactive waste in Kansas

    International Nuclear Information System (INIS)

    Bayne, C.K.; Brinkley, C.

    1972-09-01

    The three sites selected by the AEC for additional study for the disposal of radioactive wastes in Kansas are; Site A located in south-central Lincoln County, Site D-2 located in south-central Wichita County, and Site A-1 located in north-western Lincoln County. Results of the study show that all sites failed to meet the detailed criteria. Areas A and A-1 fail to meet the criteria concerning thickness and quality. Area D-2 fails to meet the criteria concerning quality and mineability of the salt. Areas west of Site A-1 and in south-central Harper County, in the authors' opinion, appear to be the best prospects for future study in Kansas

  4. Cementation of liquid radioactive waste with high content of borate salts

    International Nuclear Information System (INIS)

    Gorbunova, O.

    2015-01-01

    The report reviews the ways of optimization of cementation of boron-containing liquid radioactive waste. The most common way to hardening the low-level liquid radioactive waste (LRW) is the cementation. However, boron-containing liquid radioactive waste with low pH values cannot be cemented without alkaline additives, to neutralize acid forms of borate compounds. Cement setting without additives happens only on 14-56 days, the compounds have low strength, and hence an insufficient reliability of radionuclides fixation in the cement matrix. The alkaline additives increase the volume of the final cement compound which enhances financial and operational costs. In order to control the speed of hardening of cement solution with a boron-containing liquid radioactive waste and to remove the components that prevent hardening of cement solution, it is proposed an electromagnetic treatment of LRW in the vortex layer of ferromagnetic particles. The results of infrared spectroscopy show, that electromagnetic treatment of liquid radioactive waste changes the ionic forms of the borates and raises the pH due to the dissociation of the oxygen and hydrogen bonds in the aqueous solutions of the boron compounds. The various types of ferromagnetic activators of the vortex layer have been investigated, including the highly dispersed nano-powders and the magnetic phases of the iron oxides. It has been determined the technological parameters of the electromagnetic treatment of liquid radioactive waste and the subsequent cementation of this type of LRW. By using the method of scanning electron microscopy it has been shown, that the nano-particles of magnetic phases of the ferric oxides are involved in phase formation of hydro-aluminum-calcium ferrites in the early stages of hardening and improving strength of the cement compounds with liquid radioactive waste. (authors)

  5. Calculations on the development in space and time of the temperature field around a repository of medium and high active wastes in a salt formation

    International Nuclear Information System (INIS)

    Delisle, G.

    1980-01-01

    The concept of nuclear waste disposal of th of the Federal Republic of Germany calls for the burial of the wastes within a salt formation. A small portion of the wastes will generate heat after the disposal procedure. A temperature rise within the salt formation, in space and time limited, will be the consequence. The temperature change at any point in the near or far field of the disporal area can be calculated with the aid of numerical models. The thermal parameters representative for the bulk material of the Zechstein formation in NW-Germany, on which the calculations are based, will be discussed in detail. The interrelation between the concentration of heat producing wastes in the disposal field and the maximum average temperature in the salt formation will be treated. By defining numerical models, which are based on assumed shapes of a salt dome and a disposal area, the temperature development in the near and far field of a nuclear repository are shown. (orig.) [de

  6. Geoprospective study of a nuclear waste repository: salt domes; Bibliographic study of their genesis

    International Nuclear Information System (INIS)

    Billaux, D.; Robelin, C.

    1985-01-01

    This report appraises, from the results of a bibliographical study, the possibility of beginning of a domal rise from the salt layer in which or above which would have been placed a repository. The physical mecanisms of salt creep are first screened, together with the factors determining their intensity and relative importance. These factors are primarily the temperature and the state of stress. Semi empirical laws are given for some mecanisms. Present knowledge about the state of the salt in the ground are then examined: we are not able to satisfactorily calculate ''in situ'' stresses, or to explain the existence of an important shear stress, that has been pointed out by most of the stress measurements. The retrospective study of the genesis of existing domes brings an insight into their correlation with sedimentary and tectonic phases. Model studies help us to interpret the distances between domes, and to explain the scale of this phenomenon. After recapitulating the various factors of some importance, we find that the probability of a dome rise from a previously static layer is low, in the time lap we are interested in (100 000 years). Such a rise would have to be triggered by important changes in the sedimentation, erosion or tectonic activity on the site

  7. In situ investigations on the impact of heat production and gamma radiation with regard to high-level radioactive waste disposal in rock salt formations

    International Nuclear Information System (INIS)

    Rothfuchs, T.

    1986-01-01

    Deep geological formations especially rock salt formations, are considered worldwide as suitable media for the final disposal of radioactive high-level waste (HLW). In the Federal Republic of Germany, the Institut fur Tieflagerung of the Gesellschaft fur Strahlen- und Umweltforschung mbH Munchen operates the Asse Salt Mine as a pilot facility for testing the behavior of an underground nuclear waste repository. The tests are performed using heat and radiation sources to simulate disposed HLW canisters. The measured data obtained since 1965 show that the thermomechanical response of the salt formation and the physical/chemical changes in the vicinity of disposal boreholes are not a serious concern and that their long-term consequences can be estimated based on theoretical considerations and in-situ investigations

  8. Identification of release scenarios for a repository of radioactive waste in a salt dome in the Netherlands

    International Nuclear Information System (INIS)

    Glasbergen, P.; Hamstra, J.

    1981-01-01

    A review is presented of the long-term scenarios used in the safety analysis which was carried out for the disposal of radioactive waste in salt domes in the Netherlands. The long-term analysis involved the following natural processes or events: climatological and sea-level changes, glacial erosion, diapirism, subsidence, faulting and dissolution. The model calculations which were carried out showed the dominant parameters: the rate of diapirism and the rate of subsurface dissolution of rock salt. During the operational period the intrusion of water in the repository was considered to be the most hazardous event. Because the layout of the disposal mine, the disposal geometry and the disposal mining procedures were still under consideration, the first approach of a release scenario was made on a generic basis. A generic scenario is presented for the events during the flooding of the repository. The transport ways of water through the repository and its surroundings are indicated. It is concluded that release scenario analysis for long-term periods and for the operational period provides essential information to optimize the overall disposal system in an iterative process

  9. Ultrasonic testing of a sealing construction made of salt concrete in an underground disposal facility for radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Krause, Martin; Effner, Ute Antonie; Milmann, Boris; Voelker, Christoph; Wiggenhauser, Herbert [Federal Institute for Materials Research and Testing (BAM), Berlin (Germany); Mauke, Ralf [The Federal Office for Radiation Protection, Salzgitter (Germany)

    2015-07-01

    For the closure of radioactive waste disposal facilities engineered barriers- so called ''drift seals'' are used. The purpose of these barriers is to constrain the possible infiltration of brine and to prevent the migration of radionuclides into the biosphere. In a rock salt mine a large scale in-situ experiment of a sealing construction made of salt concrete was set up to prove the technical feasibility and operability of such barriers. In order to investigate the integrity of this structure, non-destructive ultrasonic measurements were carried out. Therefore two different methods were applied at the front side of the test-barrier: 1 Reflection measurements from boreholes 2 Ultrasonic imaging by means of scanning ultrasonic echo methods This extended abstract is a short version of an article to be published in a special edition of ASCE Journal that will briefly describe the sealing construction, the application of the non-destructive ultrasonic measurement methods and their adaptation to the onsite conditions -as well as parts of the obtained results. From this a concept for the systematic investigation of possible contribution of ultrasonic methods for quality assurance of sealing structures may be deduced.

  10. Drying of residue and separation of nitrate salts in the sludge waste for the lagoon sludge treatment

    International Nuclear Information System (INIS)

    Hwang, D. S.; Lee, K. I.; Choi, Y. D.; Hwang, S. T.; Park, J. H.

    2003-01-01

    This study investigated the dissolution property of nitrate salts in the dissolution process by water and the drying property of residue after separating nitrates in a series of the processes for the sludge treatment. Desalination was carried out with the adding ratio of water and drying property was analyzed by TG/DTA, FTIR, and XRD. Nitrate salts involved in the sludge were separated over 97% at the water adding ratio of 2.5. But a small quantity of calcium and sodium nitrate remained in the residue These were decomposed over 600 .deg. C and calcium carbonate, which was consisted mainly of residue, was decomposed into calcium oxide over 750 .deg. C. The residue have to be decomposed over 800 .deg. C to converse uranyl nitrate of six value into the stable U 3 O 8 of four value. As a result of removing the nitrates at the water adding ratio of 2.5 and drying the residue over 900 .deg. C, volume of the sludge waste decreased over 80%

  11. Salton Sea Geothermal Field, California, as a near-field natural analog of a radioactive waste repository in salt

    Science.gov (United States)

    Elders, W. A.; Cohen, L. H.

    1983-11-01

    Since high concentrations of radionuclides and high temperatures are not normally encountered in salt domes or beds, finding an exact geologic analog of expected near-field conditions in a mined nuclear waste repository in salt will be difficult. The Salton Sea Geothermal Field, however, provides an opportunity to investigate the migration and retardation of naturally occurring U, Th, Ra, Cs, Sr and other elements in hot brines which have been moving through clay-rich sedimentary rocks for up to 100,000 years. The more than thirty deep wells drilled in this field to produce steam for electrical generation penetrate sedimentary rocks containing concentrated brines where temperatures reach 3650C at only 2 km depth. The brines are primarily Na, K, Ca chlorides with up to 25% of total dissolved solids; they also contain high conentrations of metals such as Fe, Mn, Li, Zn, and Pb. This report describes the geology, geophysics and geochemistry of this system as a prelude to a study of the mobility of naturally occurring radionuclides and radionuclide analogs within it.

  12. Radioactive waste isolation in salt: rationale and methodology for Argonne-conducted reviews of site characterization programs

    International Nuclear Information System (INIS)

    Harrison, W.; Ditmars, J.D.; Tisue, M.W.; Hambley, D.F.; Fenster, D.F.; Rote, D.M.

    1985-07-01

    Both regulatory and technical concerns must be addressed in Argonne-conducted peer reviews of site characterization programs for individual sites for a high-level radioactive waste repository in salt. This report describes the regulatory framework within which reviews must be conducted and presents background information on the structure and purpose of site characterization programs as found in US Nuclear Regulatory Commission (NRC) Regulatory Guide 4.17 and Title 10, Part 60, of the Code of Federal Regulations. It also presents a methodology to assist reviewers in addressing technical concerns relating to their respective areas of expertise. The methodology concentrates on elements of prime importance to the US Department of Energy's advocacy of a given salt repository system during the NRC licensing process. Instructions are given for reviewing 12 site characterization program elements, starting with performance objectives, performance issues, and levels of performance of repository subsystem components; progressing through performance assessment; and ending with plans for data acquisition and evaluation. The success of a site characterization program in resolving repository performance issues will be determined by judging the likelihood that the proposed data acquisition activities will reduce uncertainties in the performance predictions. 8 refs., 3 figs., 5 tabs

  13. Use of ferric- and ferrous-salts in liquid waste treatment processes

    International Nuclear Information System (INIS)

    Efremenkov, V.M.; Toropov, I.G.; Toropova, V.V.; Satsukevich, V.M.; Davidov, J.P.; Jabrodsky, V.N.; Prokshin, N.E.

    1995-01-01

    Treatment of spent decontamination solutions is the most complicated task in the whole problem of management of liquid radioactive waste, because quite often they have complex compositions, which makes it difficult to find for them effective and non-expensive treatment technology. New methods of treatment of such a waste is proposed based on use of specific sorption ability of ferro- and ferri-species in solution. These species are often present in solution as the by-products, and in combination with other components of decontamination solution they can be used as initial substances for synthesis of valuable sorbents directly in treating solution. Using specific compositions and conditions in solution, it is possible to make liquid waste treatment process more effective and less expensive. Particular examples of this process is presented in this work

  14. Geologic investigation of the Virgin River Valley salt deposits, Clark County, southeastern Nevada, to investigate their suitability for possible storage of radioactive waste material as of September 1977

    International Nuclear Information System (INIS)

    1977-01-01

    The results from a geologic investigation of the Virgin River Valley salt deposits, Clark County, southeastern Nevada, to examine their suitability for further study and consideration in connection with the possible storage of radioactive waste material are given. The results indicate that (1) approximately one-half of the salt body underlies the Overton Arm of Lake Mead and that the dry land portion of the salt body that has a thickness of 1,000 feet or more covers an area of about four and one-half square miles; (2) current tectonic activity in the area of the salt deposits is believed to be confined to seismic events associated with crustal adjustments following the filling of Lake Mead; (3) detailed information on the hydrology of the salt deposit area is not available at present but it is reported that a groundwater study by the U.S. Geological Survey is now in progress; (4) there is no evidence of exploitable minerals in the salt deposit area other than evaporites such as salt, gypsum, and possibly sand and gravel; (5) the salt deposit area is located inside the Lake Mead Recreation Area, outlined on the accompanying Location Plat, and several Federal, State, and Local agencies share regulatory responsibilities for the activities in the area; (6) other salt deposit areas of Arizona and Nevada, such as the Detrital Valley, Red Lake Dome, Luke Dome, and Mormon Mesa area, and several playa lake areas of central Nevada may merit further study; and (7) additional information, as outlined, is needed to more thoroughly evaluate the salt deposits of the Virgin River Valley and other areas referred to above

  15. Problems and risks involved in the projected storage of radioactive waste in a salt dome in the northwest of the FRG

    International Nuclear Information System (INIS)

    Mauthe, F.

    1979-01-01

    Current planning envisages long-term intermediate storage of radioactive waste and the exploration of the Gorleben salt dome by deep drilling in order to start appropriate mining work in case of favourable drilling results. The statements presented here on the problem of the 'Feasibility of ultimate storage of radioactive waste in salt deposits' (subject selected by the Government of the land Lower-Saxony) are aimed at informing the general public about the difficulties and problems involved in this waste disposal project and critically assess the arguments put forward by industry and licensing authorities in order to gain acceptance for this politically delicate project; the argumentation discussed here mainly refers to the field of geological science. (orig.) [de

  16. Waste isolation facility description for the spent fuel cycle, bedded salt

    International Nuclear Information System (INIS)

    1977-05-01

    Details are given on surface facilities, shafts and hoists, mine facilities, ventilation systems, land improvements, and utilities. Accidents, confinement, and safety criteria are covered. Appendices are provided on mine layout and development, mine operations, shaft construction information, and analysis concerning canister rupture inside the proposed waste isolation facility

  17. Glossary of terms used in the disposal of high-level wastes: Salt Repository Project

    International Nuclear Information System (INIS)

    1987-02-01

    This glossary provides definitions of words and phrases specific to, or used in a special way in, documents of the US Department of Energy's Civilian Radioactive Waste Management Program. In many cases, two or more definitions of a word or phrase are given. Sources are provided for all definitions. 33 refs

  18. Expedited demonstration of molten salt mixed waste treatment technology. Addendum 1

    International Nuclear Information System (INIS)

    Holtz, E.H. von; Hopper, R.W.; Adamson, M.G.

    1995-01-01

    The Final Forms portion (Section 4) of the TTP SF-2410-03 final report was incomplete. This was noted under the subsection ''Task Variances.'' The present report documents the work that was unfinished at that time, arranged in accord with the subsections of the Final Report. An assessment of the overall immobilization efficacy of polymer microencapsulation, as supported by this study, has been added. The study and demonstration of the polyethylene microencapsulation of salt residues is continuing under other auspices. A stand-alone report combining the results of the continuation with the contents of this memorandum and of Section 4 of the Final Report will be issued in later this year

  19. National waste terminal storage repository in a bedded salt formation for spent unreprocessed fuel. Quality assurance program for licensing

    International Nuclear Information System (INIS)

    1978-12-01

    A National Waste Terminal Storage Repository, in bedded salt, for spent unreprocessed fuel is the subject of a conceptual design project which began in January 1977. This volume presents a preliminary quality assurance program to guide the license applicant in developing a detailed program that will be compatible with anticipated National Waste Terminal Storage (NWTSR2) contracting arrangements and provide the documentation required by regulatory bodies. This QA program is designed to provide confidence that the quality-related activities pertaining to safety-related structures, systems, and components will be identified and controlled. Specific responsibilities for quality-related activities are documented and assigned to personnel and organizations for the major phases of facility design and construction. These responsibilities encompass a broad range of activities and are addressed in this preliminary program. The quality assurance program elements are organized and discussed herein as follows: (1) quality assurance during design and construction; (2) the applicant (DOE); (3) siting contractor; (4) architect/engineer; (5) project field management; and (6) operations contractor

  20. Radioactive waste isolation in salt: peer review of the Office of Nuclear Waste Isolation's Geochemical Program Plan

    International Nuclear Information System (INIS)

    Harrison, W.; Seitz, M.; Fenster, D.; Lerman, A.; Brookins, D.; Tisue, M.

    1984-02-01

    Describe the management program for coordinating subcontractors and their work, and integrating research results. Appropriate flowcharts should be included. Provide more information on the overall scope of the program. For each subcontractor, provide specific workscopes that indicate whether analytical activities are developmental or routine, approximate number of analyses to be made, and something of the adequacy of the analyses to meet program goals. Indicate interfaces with other earth-science disciplines like hydrology and with other groups doing relevant geochemical research and engineering design. Address the priorities for each activity or group of activities. High priority should be given to early development of a geochemical statement of what constitutes suitable salt for a repository. Reference standard procedures for sampling, sample preservation, and sample analysis wherever appropriate or, if not appropriate, indicate that any deviations from standard procedures will be documented. Ensure that appropriate quality assurance procedures will be followed for the procedures listed above. Include specific procedures for the choice, verification, validation, and documentation of computer codes related to the geochemical aspects of repository performance assessment. Include activities addressing regional hydrochemistry and make clear that each principal hydrogeologic unit at each site will be studied geochemically. Indicate that proposed plans for obtaining hydrogeochemical data will be included in each site characterization plan. Describe how site geochemical stability will be handled, especially with respect to dissolution, postemplacement geochemistry, human influences, and climatic variations. Minor recommendations and suggested improvements in the text of the plan are given in Sec. 5

  1. Peroxisome Proliferator-Activated Receptor-α Activation Decreases Mean Arterial Pressure, Plasma Interleukin-6, and COX-2 While Increasing Renal CYP4A Expression in an Acute Model of DOCA-Salt Hypertension

    Directory of Open Access Journals (Sweden)

    Dexter L. Lee

    2011-01-01

    Full Text Available Peroxisome proliferator-activated receptor-alpha (PPAR-α activation by fenofibrate reduces blood pressure and sodium retention during DOCA-salt hypertension. PPAR-α activation reduces the expression of inflammatory cytokines, such as interleukin-6 (IL-6. Fenofibrate also induces cytochrome P450 4A (CYP4A and increases 20-hydroxyeicosatetraenoic acid (20-HETE production. This study tested whether the administration of fenofibrate would reduce blood pressure by attenuating plasma IL-6 and renal expression of cyclooxygenase-2 (COX-2, while increasing expression of renal CYP4A during 7 days of DOCA-salt hypertension. We performed uni-nephrectomy on 12–14 week old male Swiss Webster mice and implanted biotelemetry devices in control, DOCA-salt (1.5 mg/g treated mice with or without fenofibrate (500 mg/kg/day in corn oil, intragastrically. Fenofibrate significantly decreased mean arterial pressure and plasma IL-6. In kidney homogenates, fenofibrate increased CYP4A and decreased COX-2 expression. There were no differences in renal cytochrome P450, family 2, subfamily c, polypeptide 23 (CYP2C23 and soluble expoxide hydrolase (sEH expression between the groups. Our results suggest that the blood pressure lowering effect of PPAR-α activation by fenofibrate involves the reduction of plasma IL-6 and COX-2, while increasing CYP4A expression during DOCA-salt hypertension. Our results may also suggest that PPAR-α activation protects the kidney against renal injury via decreased COX-2 expression.

  2. Summary Report of Laboratory Testing to Establish the Effectiveness of Proposed Treatment Methods for Unremediated and Remediated Nitrate Salt Waste Streams

    Energy Technology Data Exchange (ETDEWEB)

    Anast, Kurt Roy [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Funk, David John [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-05-12

    The inadvertent creation of transuranic waste carrying hazardous waste codes D001 and D002 requires the treatment of the material to eliminate the hazardous characteristics and allow its eventual shipment and disposal at the Waste Isolation Pilot Plant (WIPP). This report documents the effectiveness of two treatment methods proposed to stabilize both the unremediated and remediated nitrate salt waste streams (UNS and RNS, respectively). The two technologies include the addition of zeolite (with and without the addition of water as a processing aid) and cementation. Surrogates were developed to evaluate both the solid and liquid fractions expected from parent waste containers, and both the solid and liquid fractions were tested. Both technologies are shown to be effective at eliminating the characteristic of ignitability (D001), and the addition of zeolite was determined to be effective at eliminating corrosivity (D002), with the preferred option1 of zeolite addition currently planned for implementation at the Waste Characterization, Reduction, and Repackaging Facility. During the course of this work, we established the need to evaluate and demonstrate the effectiveness of the proposed remedy for debris material, if required. The evaluation determined that Wypalls absorbed with saturated nitrate salt solutions exhibit the ignitability characteristic (all other expected debris is not classified as ignitable). Follow-on studies will be developed to demonstrate the effectiveness of stabilization for ignitable Wypall debris. Finally, liquid surrogates containing saturated nitrate salts did not exhibit the characteristic of ignitability in their pure form (those neutralized with Kolorsafe and mixed with sWheat did exhibit D001). As a result, additional nitrate salt solutions (those exhibiting the oxidizer characteristic) will be tested to demonstrate the effectiveness of the remedy.

  3. Radioactive waste isolation in salt: peer review of the Office of Nuclear Waste Isolation's reports on multifactor life testing of waste package materials

    International Nuclear Information System (INIS)

    McPheeters, C.C.; Harrison, W.; Ditmars, J.D.; Lerman, A.; Rote, D.M.; Edgar, D.E.; Hambley, D.F.

    1984-09-01

    Two documents that provide the approaches in designing a test program to investigate uniform corrosion of low-carbon cash steel in a salt repository environment were reviewed. Recommendations are made by the Peer Review Panel for improving the two reports

  4. Properties of salt-saturated concrete and grout after six years in situ at the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    Wakeley, L.D.; Harrington, P.T.; Weiss, C.A. Jr.

    1993-06-01

    Samples of concrete and grout were recovered from short boreholes in the repository floor at the Waste Isolation Pilot Plant more than six years after the concrete and grout were placed. Plugs from the Plug Test Matrix of the Plugging and Sealing Program of Sandia National Laboratories were overcored to include a shell of host rock. The cores were analyzed at the Waterways Experiment Station to assess their condition after six years of service, having potentially been exposed to those aspects of their service environment (salt, brine, fracturing, anhydrite, etc.) that could cause deterioration. Measured values of compressive strength and pulse velocity of both the grout and the concrete equaled or exceeded values from tests performed on laboratory-tested samples of the same mixtures at ages of one month to one year after casting. The phase assemblages had changed very little. Materials performed as intended and showed virtually no chemical or physical evidence of deterioration. The lowest values for strength and pulse velocity were measured for samples taken from the Disturbed Rock Zone, indicating the influence of cracking in this zone on the properties of enclosed seal materials. There was evidence of movement of brine in the system. Crystalline phases containing magnesium, potassium, sulfate, and other ions had been deposited on free surfaces in fractures and pilot holes. There was a reaction rim in the anhydrite immediately surrounding each recovered borehole plug, suggesting interaction between grout or concrete and host rock. However, the chemical changes apparent in this reaction rim were not reflected in the chemical composition of the adjacent concrete or grout. The grout and concrete studied here showed no signs of the deterioration found to have occurred in some parts of the concrete liner of the Waste Isolation Pilot Plant waste handling shaft

  5. Methods and results of the investigation of the thermomechanical behaviour of rock salt with regard to the final disposal of high-level radioactive wastes

    International Nuclear Information System (INIS)

    Wieczorek, K.; Klarr, K.

    1993-01-01

    This report summarizes the knowledge about thermal and mechanical behaviour of rock salt that has been accumulated by various R and D institutions in Germany from laboratory and in situ investigations. An important objective is to give a comprehensive overview of the investigation methods and instruments available and to discuss these methods and instruments with regard to their applicability and reliability for the investigation of the thermomechanical effects of high level radioactive waste emplacement in rock salt formations. The report is focused on the activities of the GSF-Institut fur Tieflagerung in the Asse mine regarding the disposal of high and intermediate level radioactive waste during the last decades. The design and the results of the most important in situ experiments are presented and discussed in detail. The results are compared to model calculations in order to evaluate the reliability of both the measurements and the calculation results. The relevance of the results for the situation in Spain is discussed in a separate chapter. As the investigations in Germany have been performed in domal salt, while the Spanish concept is based on waste disposal in bedded salt, significant differences in the thermomechanical behaviour cannot be excluded. The investigation methods, however, will be applicable. (Author)

  6. A Dual-Continuum Model for Brine Migration in Salt Associated with Heat-Generating Nuclear Waste: Fully Coupled Thermal-Hydro-Mechanical Analysis

    Science.gov (United States)

    Hu, M.; Rutqvist, J.

    2017-12-01

    The disposal of heat-generating nuclear waste in salt host rock establishes a thermal gradient around the waste package that may cause brine inclusions in the salt grains to migrate toward the waste package. In this study, a dual-continuum model is developed to analyze such a phenomenon. This model is based on the Finite Volume Method (FVM), and it is fully thermal-hydro-mechanical (THM) coupled. For fluid flow, the dual-continuum model considers flow in the interconnected pore space and also in the salt grains. The mass balance of salt and water in these two continua is separately established, and their coupling is represented by flux associated with brine migration. Together with energy balance, such a system produces a coupled TH model with strongly nonlinear features. For mechanical analysis, a new formulation is developed based on the Voronoi tessellated mesh. By relating each cell to several connected triangles, first-order approximation is constructed. The coupling between thermal and mechanical fields is only considered in terms of thermal expansion. And the coupling between the hydraulic and mechanical fields in terms of pore-volume effects is consistent with Biot's theory. Therefore, a fully coupled THM model is developed. Several demonstration examples are provided to verify the model. Last the new model is applied to analyze coupled THM behavior and the results are compared with experimental data.

  7. Stabilization/Solidification of radioactive molten salt waste by using xSiO2-yAl2O3-zP2O5 material

    International Nuclear Information System (INIS)

    Hwan-Seo Park; In-Tae Kim; Yong-Zun Cho; Seong-Won Park; Eung-Ho Kim

    2008-01-01

    Molten salt waste generated from the electro metallurgical process to recover uranium and transuranic elements is considered as one of problematic wastes to be difficult to immobilize into a durable for final disposal. As an alternative, this study suggested a new method performed at molten state, where dechlorination was achieved with a new inorganic material containing SiO 2 , Al 2 O 3 and P 2 O 5 (SAP). The SAP as a reactive material to molten salt was prepared by a conventional sol-gel process. The prepared SAPs were reacted with each metal chloride, LiCl, CsCl, SrCl 2 and CeCl 3 at 650 deg. C for 6 hours and also were reacted with simulated salt waste consisting of 90 wt% LiCl, 6.8 wt% CsCl and 3.2 wt% SrCl 2 at different waste loading. All the reactions were carried out in oxidative atmosphere and metal chlorides were effectively converted into stable products under a reasonable reaction ratio

  8. Effects of combustion and operating conditions on PCDD/PCDF emissions from power boilers burning salt-laden wood waste.

    Science.gov (United States)

    Leclerc, Denys; Duo, Wen Li; Vessey, Michelle

    2006-04-01

    This paper discusses the effects of combustion conditions on PCDD/PCDF emissions from pulp and paper power boilers burning salt-laden wood waste. We found no correlation between PCDD/PCDF emissions and carbon monoxide emissions. A good correlation was, however, observed between PCDD/PCDF emissions and the concentration of stack polynuclear aromatic hydrocarbons (PAHs) in the absence of TDF addition. Thus, poor combustion conditions responsible for the formation of products of incomplete combustion (PICs), such as PAHs and PCDD/PCDF precursors, increase PCDD/PCDF emissions. PAH concentrations increased with higher boiler load and/or low oxygen concentrations at the boiler exit, probably because of lower available residence times and insufficient excess air. Our findings are consistent with the current understanding that high ash carbon content generally favours heterogeneous reactions leading to either de novo synthesis of PCDD/PCDFs or their direct formation from precursors. We also found that, in grate-fired boilers, a linear increase in the grate/lower furnace temperature produces an exponential decrease in PCDD/PCDF emissions. Although the extent of this effect appears to be mill-specific, particularly at low temperatures, the results indicate that increasing the combustion temperature may decrease PCDD/PCDF emissions. It must be noted, however, that there are other variables, such as elevated ESP and stack temperatures, a high hog salt content, the presence of large amounts of PICs and a high Cl/S ratio, which contribute to higher PCDD/PCDFs emissions. Therefore, higher combustion temperatures, by themselves, will not necessarily result in low PCDD/PCDFs emissions.

  9. Effects of resource activities upon repository siting and waste containment with reference to bedded salt

    International Nuclear Information System (INIS)

    Ashby, J.; Rowe, J.

    1980-02-01

    The primary consideration for the suitability of a nuclear waste repository site is the overall ability of the repository to safely contain radioactive waste. This report is a discussion of the past, present, and future effects of resource activities on waste containment. Past and present resource activities which provide release pathways (i.e., leaky boreholes, adjacent mines) will receive initial evaluation during the early stages of any repository site study. However, other resource activities which may have subtle effects on containment (e.g., long-term pumping causing increased groundwater gradients, invasion of saline water causing lower retardation) and all potential future resource activities must also be considered during the site evaluation process. Resource activities will affect both the siting and the designing of repositories. Ideally, sites should be located in areas of low resource activity and low potential for future activity, and repository design should seek to eliminate or minimize the adverse effects of any resource activity. Buffer zones should be created to provide areas in which resource activities that might adversely affect containment can be restricted or curtailed. This could mean removing large areas of land from resource development. The impact of these frozen assets should be assessed in terms of their economic value and of their effect upon resource reserves. This step could require a major effort in data acquisition and analysis followed by extensive numerical modeling of regional fluid flow and mass transport. Numerical models should be used to assess the effects of resource activity upon containment and should include the cumulative effects of different resource activities. Analysis by other methods is probably not possible except for relatively simple cases

  10. Thermal Analysis of Disposal of High-Level Nuclear Waste in a Generic Bedded Salt repository using the Semi-Analytical Method.

    Energy Technology Data Exchange (ETDEWEB)

    Hadgu, Teklu [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Matteo, Edward N. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-05-01

    An example case is presented for testing analytical thermal models. The example case represents thermal analysis of a generic repository in bedded salt at 500 m depth. The analysis is part of the study reported in Matteo et al. (2016). Ambient average ground surface temperature of 15°C, and a natural geothermal gradient of 25°C/km, were assumed to calculate temperature at the near field. For generic salt repository concept crushed salt backfill is assumed. For the semi-analytical analysis crushed salt thermal conductivity of 0.57 W/m-K was used. With time the crushed salt is expected to consolidate into intact salt. In this study a backfill thermal conductivity of 3.2 W/m-K (same as intact) is used for sensitivity analysis. Decay heat data for SRS glass is given in Table 1. The rest of the parameter values are shown below. Results of peak temperatures at the waste package surface are given in Table 2.

  11. Summary Report of Comprehensive Laboratory Testing to Establish the Effectiveness of Proposed Treatment Methods for Unremediated and Remediated Nitrate Salt Waste Streams

    Energy Technology Data Exchange (ETDEWEB)

    Anast, Kurt Roy [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Funk, David John [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hargis, Kenneth Marshall [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-10-04

    The inadvertent creation of transuranic waste carrying hazardous waste codes D001 and D002 requires the treatment of the material to eliminate the hazardous characteristics and allow its eventual shipment and disposal at the Waste Isolation Pilot Plant (WIPP). This report documents the effectiveness of two treatment methods proposed to stabilize both the unremediated and remediated nitrate salt waste streams (UNS and RNS, respectively) at Los Alamos National Laboratory (LANL). The two technologies include the addition of zeolite (with and without the addition of water as a processing aid) and cementation. Surrogates were developed to evaluate both the solid and liquid fractions expected from parent waste containers, and both the solid and liquid fractions were tested. Both technologies are shown to be effective at eliminating the characteristic of ignitability (D001), and the addition of zeolite was determined to be effective at eliminating corrosivity (D002), with the preferred option1 of adding zeolite currently planned for implementation at LANL’s Waste Characterization, Reduction, and Repackaging Facility (WCRRF). The course of this work verified the need to evaluate and demonstrate the effectiveness of the proposed remedy for debris material, if required. The evaluation determined that WypAlls, cheesecloth, and Celotex absorbed with saturated nitrate salt solutions exhibit the ignitability characteristic (all other expected debris is not classified as ignitable). Finally, liquid surrogates containing saturated nitrate salts did not exhibit the characteristic of ignitability in their pure form (those neutralized with Kolorsafe and mixed with sWheat did exhibit D001). Sensitivity testing and an analysis were conducted to evaluate the waste form for reactivity. Tests included subjecting surrogate material to mechanical impact, friction, electrostatic discharge and thermal insults. The testing confirmed that the waste does not exhibit the characteristic of

  12. Effect of biosolid waste compost on soil respiration in salt-affected soils

    Science.gov (United States)

    Raya, Silvia; Gómez, Ignacio; García, Fuensanta; Navarro, José; Jordán, Manuel Miguel; Belén Almendro, María; Martín Soriano, José

    2013-04-01

    A great part of mediterranean soils are affected by salinization. This is an important problem in semiarid areas increased by the use of low quality waters, the induced salinization due to high phreatic levels and adverse climatology. Salinization affects 25% of irrigated agriculture, producing important losses on the crops. In this situation, the application of organic matter to the soil is one of the possible solutions to improve their quality. The main objective of this research was to asses the relation between the salinity level (electrical conductivity, EC) in the soil and the response of microbial activity (soil respiration rate) after compost addition. The study was conducted for a year. Soil samples were collected near to an agricultural area in Crevillente and Elche, "El Hondo" Natural Park (Comunidad de Regantes from San Felipe Neri). The experiment was developed to determine and quantify the soil respiration rate in 8 different soils differing in salinity. The assay was done in close pots -in greenhouse conditions- containing soil mixed with different doses of sewage sludge compost (2, 4 and 6%) besides the control. They were maintained at 60% of water holding capacity (WHC). Soil samples were analyzed every four months for a year. The equipment used to estimate the soil respiration was a Bac-Trac and CO2 emitted by the soil biota was measured and quantified by electrical impedance changes. It was observed that the respiration rate increases as the proportion of compost added to each sample increases as well. The EC was incremented in each sampling period from the beginning of the experiment, probably due to the fact that soils were in pots and lixiviation was prevented, so the salts couldńt be lost from soil. Over time the compost has been degraded and, it was more susceptible to be mineralized. Salts were accumulated in the soil. Also it was observed a decrease of microbial activity with the increase of salinity in the soil. Keywords: soil

  13. [Muscle-wasting in end stage renal disease in dialysis treatment: a review].

    Science.gov (United States)

    Battaglia, Yuri; Galeano, Dario; Cojocaru, Elena; Fiorini, Fulvio; Forcellini, Silvia; Zanoli, Luca; Storari, Alda; Granata, Antonio

    2016-01-01

    Progressive and generalized loss of muscle mass (muscle wasting) is a frequent complication in dialysis patients. Common uremic signs and symptoms such as insulin-resistance, increase in glucocorticoid activity, metabolic acidosis, malnutrition, inflammation and dialysis per se contribute to muscle wasting by modulating proteolytic intracellular mechanisms (ubiquitin-proteasome system, activation of caspase-3 and IGF-1/PI3K/Akt pathway). Since muscle wasting is associated with an increase in mortality, bone fractures and worsening in life quality, a prompt and personalised diagnostic and therapeutic approach seems to be essential in dialysis patients. At present, nuclear magnetic resonance (NMR), computed tomography (CT), dual-energy x-ray absorptiometry (DXA), impedance analysis, bioelectric impedance analysis (BIA) and anthropometric measurements are the main tools used to assess skeletal muscle mass. Aerobic and anaerobic training programmes and treatment of uremic complications reduce muscle wasting and increase muscle strength in uremic patients. The present review analyses the most recent data about the physiopathology, diagnosis, therapy and future perspectives of treatment of muscle wasting in dialysis patients.

  14. Summary of four release consequence analyses for hypothetical nuclear waste repositories in salt and granite

    International Nuclear Information System (INIS)

    Cole, C.R.; Bond, F.W.

    1980-12-01

    Release consequence methology developed under the Assessment of Effectiveness of Geologic Isolation Systems (AEGIS) program has now been applied to four hypothetical repository sites. This paper summarizes the results of these four studies in order to demonstrate that the far-field methodology developed under the AEGIS program offers a practical approach to the post-closure safety assessment of nuclear waste repositories sited in deep continental geologic formations. The four studies are briefly described and compared according to the following general categories: physical description of the repository (size, inventory, emplacement depth); geologic and hydrologic description of the site and the conceptual hydrologic model for the site; description of release scenario; hydrologic model implementation and results; engineered barriers and leach rate modeling; transport model implementation and results; and dose model implementation and results. These studies indicate the following: numerical modeling is a practical approach to post-closure safety assessment analysis for nuclear waste repositories; near-field modeling capability needs improvement to permit assessment of the consequences of human intrusion and pumping well scenarios; engineered barrier systems can be useful in mitigating consequences for postulated release scenarios that short-circuit the geohydrologic system; geohydrologic systems separating a repository from the natural biosphere discharge sites act to mitigate the consequences of postulated breaches in containment; and engineered barriers of types other than the containment or absorptive type may be useful

  15. Molten salt breeder reactor

    International Nuclear Information System (INIS)

    1977-01-01

    MSBR Study Group formed in October 1974 has studied molten salt breeder reactor and its various aspects. Usage of a molten salt fuel, extremely interesting as reactor chemistry, is a great feature to MSBR; there is no need for separate fuel making, reprocessing, waste storage facilities. The group studied the following, and these results are presented: molten salt technology, molten salt fuel chemistry and reprocessing, reactor characteristics, economy, reactor structural materials, etc. (Mori, K.)

  16. Waste form evaluation for RECl 3 and REO x fission products separated from used electrochemical salt

    Energy Technology Data Exchange (ETDEWEB)

    Riley, Brian J.; Pierce, David A.; Crum, Jarrod V.; Williams, Benjamin D.; Snyder, Michelle M. V.; Peterson, Jacob A.

    2018-04-01

    The work presented here is based off the concept that the rare earth chloride (RECl3) fission products mixture within the used electrorefiner (ER) salt can be selectively removed as RECl3 (not yet demonstrated) or precipitated out as REOCl through oxygen sparging (has been demonstrated). This paper presents data showing the feasibility of immobilizing a mixture of RECl3’s at 10 mass% into a TeO2-PbO glass and it shows that this same mixture of RECl3’s can be oxidized to REOCl at 300°C and then to REOx by 1200°C. When the REOx mixture is heated at temperatures >1200°C, the ratios of REOx’s change. The mixture of REOx was then immobilized in a LABS glass at a high loading of 60 mass%. Both the TeO2-PbO glass and LABS glass systems show good chemical durability. The advantages and disadvantages of tellurite and LABS glasses are compared.

  17. Site characterization plan conceptual design report for a high-level nuclear waste repository in salt, vertical emplacement mode: Volume 1

    International Nuclear Information System (INIS)

    1987-12-01

    This Conceptual Design Report describes the conceptual design of a high-level nuclear waste repository in salt at a proposed site in Deaf Smith County, Texas. Waste receipt, processing, packing, and other surface facility operations are described. Operations in the shafts underground are described, including waste hoisting, transfer, and vertical emplacement. This report specifically addresses the vertical emplacement mode, the reference design for the repository. Waste retrieval capability is described. The report includes a description of the layout of the surface, shafts, and underground. Major equipment items are identified. The report includes plans for decommissioning and sealing of the facility. The report discusses how the repository will satisfy performance objectives. Chapters are included on basis for design, design analyses, and data requirements for completion of future design efforts. 105 figs., 52 tabs

  18. Site characterization plan conceptual design report for a high-level nuclear waste repository in salt, vertical emplacement mode: Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    1987-12-01

    This Conceptual Design Report describes the conceptual design of a high-level nuclear waste repository in salt at a proposed site in Deaf Smith County, Texas. Waste receipt, processing, packing, and other surface facility operations are described. Operations in the shafts underground are described, including waste hoisting, transfer, and vertical emplacement. This report specifically addresses the vertical emplacement mode, the reference design for the repository. Waste retrieval capability is described. The report includes a description of the layout of the surface, shafts, and underground. Major equipment items are identified. The report includes plans for decommissioning and sealing of the facility. The report discusses how the repository will satisfy performance objectives. Chapters are included on basis for design, design analyses, and data requirements for completion of future design efforts. 105 figs., 52 tabs.

  19. Repository environmental parameters and models/methodologies relevant to assessing the performance of high-level waste packages in basalt, tuff, and salt

    Energy Technology Data Exchange (ETDEWEB)

    Claiborne, H.C.; Croff, A.G.; Griess, J.C.; Smith, F.J.

    1987-09-01

    This document provides specifications for models/methodologies that could be employed in determining postclosure repository environmental parameters relevant to the performance of high-level waste packages for the Basalt Waste Isolation Project (BWIP) at Richland, Washington, the tuff at Yucca Mountain by the Nevada Test Site, and the bedded salt in Deaf Smith County, Texas. Guidance is provided on the identify of the relevant repository environmental parameters; the models/methodologies employed to determine the parameters, and the input data base for the models/methodologies. Supporting studies included are an analysis of potential waste package failure modes leading to identification of the relevant repository environmental parameters, an evaluation of the credible range of the repository environmental parameters, and a summary of the review of existing models/methodologies currently employed in determining repository environmental parameters relevant to waste package performance. 327 refs., 26 figs., 19 tabs.

  20. Repository environmental parameters and models/methodologies relevant to assessing the performance of high-level waste packages in basalt, tuff, and salt

    International Nuclear Information System (INIS)

    Claiborne, H.C.; Croff, A.G.; Griess, J.C.; Smith, F.J.

    1987-09-01

    This document provides specifications for models/methodologies that could be employed in determining postclosure repository environmental parameters relevant to the performance of high-level waste packages for the Basalt Waste Isolation Project (BWIP) at Richland, Washington, the tuff at Yucca Mountain by the Nevada Test Site, and the bedded salt in Deaf Smith County, Texas. Guidance is provided on the identify of the relevant repository environmental parameters; the models/methodologies employed to determine the parameters, and the input data base for the models/methodologies. Supporting studies included are an analysis of potential waste package failure modes leading to identification of the relevant repository environmental parameters, an evaluation of the credible range of the repository environmental parameters, and a summary of the review of existing models/methodologies currently employed in determining repository environmental parameters relevant to waste package performance. 327 refs., 26 figs., 19 tabs

  1. Site characterization plan conceptual design report for a high-level nuclear waste repository in salt, horizontal emplacment mode: Volume 1

    International Nuclear Information System (INIS)

    1987-12-01

    This Conceptual Design Report describes the conceptual design of a high-level nuclear waste repository in salt at a proposed site in Deaf Smith County, Texas. Waste receipt, processing, packaging, and other surface facility operations are described. Operations in the shafts and underground are described, including waste hoisting, transfer, and horizontal emplacement. This report specifically addresses the horizontal emplacement mode, the passive alternate design for the repository. Waste retrieval capability is described. The report includes a description of the layout of the surface, shafts, and underground. Major equipment items are identified. The report includes plans for decommissioning and sealing of the facility. The report discusses how the repository will satisfy performance objectives. Chapters are included on basis for design, design analyses, and data requirements for completion of future design efforts. 105 figs., 52 tabs

  2. Sellafield waste radionuclides in Irish sea intertidal and salt marsh sediments.

    Science.gov (United States)

    Mackenzie, A B; Scott, R D

    1993-09-01

    Low level liquid radioactive waste discharges from the Sellafield nuclear fuel reprocessing plant in north west England had generated environmental inventories of about 3 × 10(16) Bq of(137)Cs, 6.8 × 10(14) Bq of(239,240)Pu and 8.9 × 10(14) Bq of(241)Am by 1990. Most of the(239,240)Pu and(241)Am and about 10% of the(137)Cs has been retained in a deposit of fine marine sediment close to the discharge point. The quantities of radionuclides discharged annually from Sellafield decreased by two orders of magnitude from the mid-1970s to 1990 but estimated critical group internal and external exposure decreased by less than one order of magnitude over this period. This indicates that during the period of reduced discharges, radionuclides already in the environment from previous releases continued to contribute to the critical group exposure and highlights the need to understand processes controlling the environmental distribution of the radionuclides.Redistribution of the contaminated marine sediment is potentially of major significance in this context, in particular if it results in transport of radionuclides to intertidal areas, where contact with the human population is relatively likely.A review is presented of published work relating to Sellafield waste radionuclides in Irish Sea sediments. Data on temporal and spatial trends in radionuclide concentrations and activity ratios are collated from a number of sources to show that the dominant mechanism of radionuclide supply to intertidal areas is by redistribution of the contaminated marine sediment. The implications of this mechanism of supply for trends in critical group radiation exposure are considered.

  3. Conditions for the test emplacement of intermediate-level radioactive wastes in chamber 8a of the 511 m level of the Asse Salt Mine

    International Nuclear Information System (INIS)

    1984-01-01

    The Gesellschaft fuer Strahlen- und Umweltforschung mbH (GSF) emplaces intermediate-level radioactive wastes which accumulate in an activity involving the use of radioactive materials that is licensed or reported in the Federal Republic of Germany or which are stored on an interim basis by the appropriate licensing or inspection agencies in chamber 8a of the 511 m level of the Asse Salt Mine in Remlingen near Wolfenbuettel in conjunction with an engineering test program. The type and form of the intermediate-level wastes must conform to certain conditions so that there are no hazards to personnel and the repository during transfer and subsequent storage. It is therefore necessary for the radioactive wastes to be treated and packaged before delivery in such a way that they satisfy the conditions presented in this document. The GSF shall inform the companies and organizations delivering wastes about its experiences with emplacement operations. The Conditions for the Test Emplacement of Intermediate-Level Radioactive Wastes in Chamber 8a of the 511 m Level of the Asse Salt Mine must be adapted to conform to the latest state of science and the art. The GSF must therefore reserve the right to modify the conditions, allowing for an appropriate transition period

  4. Radioactive waste isolation in salt: peer review of the Office of Nuclear Waste Isolation's plan to decommission and reclaim exploratory shafts and related facilities

    International Nuclear Information System (INIS)

    Fenster, D.F.; Schubert, J.P.; Zellmer, S.D.; Harrison, W.; Simpson, D.G.; Busch, J.S.

    1984-07-01

    The following recommendations are made for improving the Office of Nuclear Waste Isolation's plan for decommissioning and reclaiming exploratory shafts and other facilities associated with site characterization: (1) Discuss more comprehensively the technical aspects of activities related to decommissioning and reclamation. More detailed information will help convince the staff of the US Nuclear Regulatory Commission and others that the activities as outlined in the plan are properly structured and that the stated goals can be achieved. (2) Address in considerably greater detail how the proposed activities will satisfy specific federal, state, and local laws and regulations. (3) State clearly the precise purpose of the plan, preferably at the beginning and under an appropriate heading. (4) Also under an appropriate heading and immediately after the section on purpose, describe the scope of the plan. The tasks covered by this plan and closely related tasks covered by other appropriate plans should be clearly differentiated. (5) Discuss the possible environmental effects of drilling the exploratory shaft, excavating drifts in salt, and drilling boreholes as part of site characterization. Mitigation activities should be designed to counter specific potential impacts. High priority should be given to minimizing groundwater contamination and restoring the surface to a condition consistent with the proposed land use following completion of characterization activities at sites not chosen for repository construction. (6) Define ambiguous technical terms, either in the text when first introduced or in an appended glossary

  5. Dechlorination and Stabilization of Molten Salt Waste by Using xSiO2-yAl2O3- zP2O5 at Melting Temperature

    International Nuclear Information System (INIS)

    Park, Hwanseo; Kim, Intae; Kim, Hwanyoung; Kim, Joonhyung

    2007-01-01

    Molten salt waste, which is generated from the pyroprocess to separate uranium and trans-uranium elements from spent nuclear fuel, has been interested to researchers in the radioactive waste management. For its final disposal, direct immobilization into a suitable host matrix or indirect solidification by other chemical routes requires the control of chlorides and its volatility since molten salt wastes mainly consist of volatile metal chlorides. Glass-bonded sodalite (Na 6 M 2 Al 6 Si 6 O 24 Cl 2 , 1-5) suggested by Argonne National Laboratory (ANL), to the present, could be a practical solution to the immobilization of this waste, where waste form can be fabricated at about 915 .deg., lower than the melting temperature of many borosilicate glasses ( -1150 .deg.). A wet dechlorination to oxides or a thermal conversion into borate glass was suggested to remove Cl from salt waste (6-7) and it seemed that the preference of radionuclides for the intended chemical conversions or immobilizations described above could be hardly accomplished or failed, except the phosphate precipitation method suggested by Volkovich and his co-workers (8). Our research group suggested a novel method to treat molten salt waste, named GRSS (Gel-Route Stabilization/Solidification) using Si-P-Al system as a gel-forming system. This showed little vaporization during high temperature process and good leach resistance on Cs and Sr. As another method, this study suggested a method to stabilize molten salt wastes by using xSiO 2 -yAl 2 O 3 - zP 2 O 5 material. GRSS method is considered as a 'reaction system' to completely convert salt waste into stable product while the inorganic material used in this study is a stabilizer for salt wastes. Using this material, this study investigated the reactivity on different metal chlorides, thermal stability, leach-resistance and etc

  6. In situ corrosion studies on selected high level waste packaging materials under simulated disposal conditions in rock salt

    International Nuclear Information System (INIS)

    Smailos, E.; Schwarzkopf, W.; Koester, R.

    1988-01-01

    In order to qualify corrosion resistant materials for high level waste (HLW) packagings acting as a long-term barrier in a rock salt repository, the corrosion behavior of preselected materials is being investigated in laboratory-scale and in-situ experiments. This work reports about in-situ corrosion experiments on unalloyed steels, Ti 99.8-Pd, Hastelloy C4, and iron-base alloys, as nodular cast iron, Ni-Resist D4 and Si-cast iron, under simulated disposal conditions. The results of the investigations can be summarized as follows: (1) all materials investigated exhibited high resistance to corrosion under the conditions prevailing in the Brine Migration Test; (2) all materials and above all the materials with passivating oxide layers such as Ti 99.8-Pd and Hastelloy C4 which may corrode selectively already in the presence of minor amounts of brine had been resistant with respect to any type of local corrosion attack; the gamma-radiation of 3 · 10 2 Gy/h did not exert an influence on the corrosion behavior of the materials

  7. Protein-energy wasting and nutritional supplementation in patients with end-stage renal disease on hemodialysis.

    Science.gov (United States)

    Sabatino, A; Regolisti, G; Karupaiah, T; Sahathevan, S; Sadu Singh, B K; Khor, B H; Salhab, N; Karavetian, M; Cupisti, A; Fiaccadori, E

    2017-06-01

    Protein-Energy Wasting (PEW) is the depletion of protein/energy stores observed in the most advanced stages of Chronic Kidney Disease (CKD). PEW is highly prevalent among patients on chronic dialysis, and is associated with adverse clinical outcomes, high morbidity/mortality rates and increased healthcare costs. This narrative review was aimed at exploring the pathophysiology of PEW in end-stage renal disease (ESRD) on hemodialysis. The main aspects of nutritional status evaluation, intervention and monitoring in this clinical setting were described, as well as the current approaches for the prevention and treatment of ESRD-related PEW. An exhaustive literature search was performed, in order to identify the relevant studies describing the epidemiology, pathogenesis, nutritional intervention and outcome of PEW in ESRD on hemodialysis. The pathogenesis of PEW is multifactorial. Loss of appetite, reduced intake of nutrients and altered lean body mass anabolism/catabolism play a key role. Nutritional approach to PEW should be based on a careful and periodic assessment of nutritional status and on timely dietary counseling. When protein and energy intakes are reduced, nutritional supplementation by means of specific oral formulations administered during the hemodialysis session may be the first-step intervention, and represents a valid nutritional approach to PEW prevention and treatment since it is easy, effective and safe. Omega-3 fatty acids and fibers, now included in commercially available preparations for renal patients, could lend relevant added value to macronutrient supplementation. When oral supplementation fails, intradialytic parenteral nutrition can be implemented in selected patients. Copyright © 2016 Elsevier Ltd and European Society for Clinical Nutrition and Metabolism. All rights reserved.

  8. Laboratory creep and mechanical tests on salt data report (1975-1996): Waste Isolation Pilot Plant (WIPP) thermal/structural interactions program

    Energy Technology Data Exchange (ETDEWEB)

    Mellegard, K.D. [RE/SPEC Inc., Rapid City, SD (United States); Munson, D.E. [Sandia National Labs., Albuquerque, NM (United States)

    1997-02-01

    The Waste Isolation Pilot Plant (WIPP), a facility located in a bedded salt formation in Carlsbad, New Mexico, is being used by the U.S. Department of Energy to demonstrate the technology for safe handling and disposal of transuranic wastes produced by defense activities in the United States. In support of that demonstration, mechanical tests on salt were conducted in the laboratory to characterize material behavior at the stresses and temperatures expected for a nuclear waste repository. Many of those laboratory test programs have been carried out in the RE/SPEC Inc. rock mechanics laboratory in Rapid City, South Dakota; the first program being authorized in 1975 followed by additional testing programs that continue to the present. All of the WIPP laboratory data generated on salt at RE/SPEC Inc. over the last 20 years is presented in this data report. A variety of test procedures were used in performance of the work including quasi-static triaxial compression tests, constant stress (creep) tests, damage recovery tests, and multiaxial creep tests. The detailed data is presented in individual plots for each specimen tested. Typically, the controlled test conditions applied to each specimen are presented in a plot followed by additional plots of the measured specimen response. Extensive tables are included to summarize the tests that were performed. Both the tables and the plots contain cross-references to the technical reports where the data were originally reported. Also included are general descriptions of laboratory facilities, equipment, and procedures used to perform the work.

  9. Method to increase the safety of a final storage site in a salt cavern filled with solidified radioactive waste with regard to unforeseen rock movements and/or water ingress into cavities of the final storage site

    International Nuclear Information System (INIS)

    Koester, R.; Rudolph, G.; Kroebel, R.

    1986-01-01

    The wastes of weak or average radio-activity (e.g. T) are stored in barrels in a salt mine. In order to prevent leaching of the waste after the ingress of water into the salt mine, the intermediate spaces between the barrels are filled with a concrete grout. This grout consists of a water/bentonite/cement mixture, to which sand may be added, and which hardens. It forms a monolithic block. (orig./PW)

  10. Method to increase the safety of a final storage site in a salt cavern filled with solidified radioactive waste with regard to unforeseen rock movements and/or water ingress into cavities of the final storage site

    International Nuclear Information System (INIS)

    Koester, R.; Rudolph, G.; Kroebel, R.

    1980-01-01

    The wastes of weak or average radio-activity (e.g. T) are stored in barrels in a salt mine. In order to prevent leaching of the waste after the ingress of water into the salt mine, the intermediate spaces between the barrels are filled with a concrete grout. This grout consists of a water/bentonite/cement mixture, to which sand may be added, and which hardens. It forms a monolithic block. (DG) [de

  11. Geologic study of the interior Salt Domes of Northeast Texas Salt-Dome basin to investigate their suitability for possible storage of radioactive waste material

    International Nuclear Information System (INIS)

    1976-05-01

    The purpose of this study was to investigate the movement and hydrologic stability of the domes, to identify the domes which appear suitable for further study and consideration, and to outline the additional information needed to evaluate these domes. The growth of the interior salt domes appears to have slowed with geologic time and to have halted altogether. The Bullard, Whitehouse, and Keechi domes probably are not subject to significant dissolution at the present time. However, caprock found at Bullard and Whitehouse indicates that salt dissolution occurred at some period during the past 50 million years since Wilcox was deposited. It is recommended that shallow water wells be drilled and tested

  12. Worth its salt?

    Science.gov (United States)

    The idea that all underground salt deposits can serve as storage sites for toxic and nuclear waste does not always hold water—literally. According to Daniel Ronen and Brian Berkowitz of Israel's Weizmann Institute of Science and Yoseph Yechieli of the Geological Survey of Israel, some buried salt layers are in fact highly conductive of liquids, suggesting that wastes buried in their confines could easily leech into groundwater and nearby soil.When drilling three wells into a 10,000-year-old salt layer near the Dead Sea, the researchers found that groundwater had seeped into the layer and had absorbed some of its salt.

  13. Ceramicrete stabilization of radioactive-salt-containing liquid waste and sludge water. Final CRADA report.

    Energy Technology Data Exchange (ETDEWEB)

    Ehst, D.; Nuclear Engineering Division

    2010-08-04

    It was found that the Ceramicrete Specimens incorporated the Streams 1 and 2 sludges with the adjusted loading about 41.6 and 31.6%, respectively, have a high solidity. The visible cracks in the matrix materials and around the anionite AV-17 granules included could not obtain. The granules mentioned above fixed by Ceramicrete matrix very strongly. Consequently, we can conclude that irradiation of Ceramecrete matrix, goes from the high radioactive elements, not result the structural degradation. Based on the chemical analysis of specimens No.462 and No.461 used it was shown that these matrix included the formation elements (P, K, Mg, O), but in the different samples their correlations are different. These ratios of the content of elements included are about {+-} 10%. This information shows a great homogeneity of matrix prepared. In the list of the elements founded, expect the matrix formation elements, we detected also Ca and Si (from the wollastonite - the necessary for Ceramicrete compound); Na, Al, S, O, Cl, Fe, Ni also have been detected in the Specimen No.642 from the waste forms: NaCl, Al(OH){sub 3}, Na{sub 2}SO{sub 4}. Fe(OH){sub 3}, nickel ferrocyanide and Ni(NO{sub 3})2. The unintelligible results also were found from analysis of an AV-17 granules, in which we obtain the great amount of K. The X-ray radiographs of the Ceramicrete specimens with loading 41.4 % of Stream 1 and 31.6% of Stream 2, respectively showed that the realization of the advance technology, created at GEOHKI, leads to formation of excellent ceramic matrix with high amount of radioactive streams up to 40% and more. Really, during the interaction with start compounds MgO and KH{sub 2}PO{sub 4} with the present of H{sub 3}BO{sub 3} and Wollastonite this process run with high speed under the controlled regimes. That fact that the Ceramicrete matrix with 30-40% of Streams 1 and 2 have a crystalline form, not amorphous matter, allows to permit that these matrix should be very stable, reliable

  14. Adaptive strategies for post-renal handling of urine in birds

    DEFF Research Database (Denmark)

    Laverty, Gary; Skadhauge, Erik

    2008-01-01

    Birds are a diverse vertebrate class in terms of diet and habitat, but they share several common physiological features, including the use of uric acid as the major nitrogenous waste product and the lack of a urinary bladder. Instead, ureteral urine refluxes from the urodeum into the more proximal...... coprodeum and portions of the hindgut (colon or rectum and ceca). This presents a potential problem in that hyperosmotic ureteral urine in contact with the permeable epithelia of these tissues would counteract renal osmotic work. This review describes and provides examples of different strategies used...... by avian species to balance renal and post-renal changes in urien composition. The strategies described include: 1. a "reptilian" mode, with moderate renal concentrating ability, but high rates of post-renal salt and water resorption; 2. the "mammalian" strategy, in which the coprodeum effectively...

  15. Long-time leaching and corrosion tests on full-scale cemented waste forms in the Asse salt mine. Sampling and analyses 2003

    International Nuclear Information System (INIS)

    Kienzler, B.; Schlieker, M.; Bauer, A.; Metz, V.; Meyer, H.

    2004-10-01

    The paper presents the follow-up of experimental findings from full-scale leach tests performed on simulated cemented waste forms for more than 20 years in salt brines and water. Measurements cover pH, density, and the composition of leachates as well as the release of radionuclides such as Cs, U and Np. Indicators for waste form corrosion and radionuclide release is Cs and NO 3 . Corrosion of cemented waste forms depends on the pore volume of the hardened cement which is correlated to the water/cement ratio. The release of radionuclides is evaluated and compared to small-scale laboratory tests. Excellent interpretation of observed concentrations is obtained for uranium and neptunium by comparison with model calculations. (orig.)

  16. Recommended new criteria for the selection of nuclear waste repository sites in Columbia River basalt and US Gulf Coast domed salt

    International Nuclear Information System (INIS)

    Steinborn, T.L.; Wagoner, J.L.; Qualheim, B.; Fitts, C.R.; Stetkar, R.E.; Turnbull, R.W.

    1980-01-01

    Screening criteria and specifications are recommended to aid in the evaluation of sites proposed for nuclear waste disposal in basalt and domed salt. The recommended new criteria proposed in this report are intended to supplement existing repository-related criteria for nuclear waste disposal. The existing criteria are contained in 10 CFR 60 sections which define siting criteria of the Nuclear Regulatory Commission (NRC), and ONWI 33(2) which defines siting criteria of the Office of Nuclear Waste Isolation (ONWI) for the Department of Energy. The specifications are conditions or parameter values that the authors recommend be applied in site acceptance evaluations. The siting concerns covered in this report include repository depth, host rock extent, seismic setting, structural and tectonic conditions, groundwater and rock geochemistry, volcanism, surface and subsurface hydrology, and socioeconomic issues, such as natural resources, land use, and population distribution

  17. Renal-Specific Silencing of TNF (Tumor Necrosis Factor) Unmasks Salt-Dependent Increases in Blood Pressure via an NKCC2A (Na+-K+-2Cl- Cotransporter Isoform A)-Dependent Mechanism.

    Science.gov (United States)

    Hao, Shoujin; Hao, Mary; Ferreri, Nicholas R

    2018-06-01

    We tested the hypothesis that TNF (tumor necrosis factor)-α produced within the kidney and acting on the renal tubular system is part of a regulatory mechanism that attenuates increases in blood pressure in response to high salt intake. Intrarenal administration of a lentivirus construct, which specifically silenced TNF in the kidney, did not affect baseline blood pressure. However, blood pressure increased significantly 1 day after mice with intrarenal silencing of TNF ingested 1% NaCl in the drinking water. The increase in blood pressure, which was continuously observed for 11 days, promptly returned to baseline levels when mice were switched from 1% NaCl to tap water. Silencing of renal TNF also increased NKCC2 (Na + -K + -2Cl - cotransporter) phosphorylation and induced a selective increase in NKCC2A (NKCC2 isoform A) mRNA accumulation in both the cortical and medullary thick ascending limb of Henle loop that was neither associated with a compensatory decrease of NKCC2F in the medulla nor NKCC2B in the cortex. The NaCl-mediated increases in blood pressure were completely absent when NKCC2A, using a lentivirus construct that did not alter expression of NKCC2F or NKCC2B, and TNF were concomitantly silenced in the kidney. Moreover, the decrease in urine volume and NaCl excretion induced by renal TNF silencing was abolished when NKCC2A was concurrently silenced, suggesting that this isoform contributes to the transition from a salt-resistant to salt-sensitive phenotype. Collectively, the data are the first to demonstrate a role for TNF produced by the kidney in the modulation of sodium homeostasis and blood pressure regulation. © 2018 American Heart Association, Inc.

  18. Chemistry of brines in salt from the Waste Isolation Pilot Plant (WIPP), southeastern New Mexico: a preliminary investigation

    International Nuclear Information System (INIS)

    Stein, C.L.; Krumhansl, J.L.

    1986-03-01

    We present here analyses of macro- and microscopic (intracrystalline) brines observed within the WIPP facility and in the surrounding halite, with interpretations regarding the origin and history of these fluids and their potential effect(s) on long-term waste storage. During excavation, several large fluid inclusions were recovered from an area of highly recrystallized halite in a thick salt bed at the repository horizon (2150 ft below ground level). In addition, 52 samples of brine ''weeps'' were collected from walls of recently excavated drifts at the same stratigraphic horizon from which the fluid inclusion samples are assumed to have been taken. Analyses of these fluids show that they differ substantially in composition from the inclusion fluids and cannot be explained by mixing of the fluid inclusion populations. Finally, holes in the facility floor that filled with brine were sampled but with no stratographic control; therefore it is not possible to interpret the compositions of these brines with any accuracy, except insofar as they resemble the weep compositions but with greater variation in both K/Mg and Na/Cl ratios. However, the Ca and SO 4 values for the floor holes are relatively close to the gypsum saturation curve, suggesting that brines filling floor holes have been modified by the presence of gypsum or anhydrite, possibly even originating in one or more of the laterally continuous anhydrite units referred to in the WIPP literature as marker beds. In conclusion, the wide compositional variety of fluids found in the WIPP workings suggest that (1) an interconnected hydrologic system which could effectively transport radonuclides away from the repository does not exist; (2) brine migration studies and experiments must consider the mobility of intergranular fluids as well as those in inclusions; and (3) near- and far-field radionuclide migration testing programs need to consider a wide range of brine compositions rather than a few reference brines

  19. Mass transport in bedded salt and salt interbeds

    International Nuclear Information System (INIS)

    Hwang, Y.; Pigford, T.H.; Chambre, P.L.; Lee, W.W.L.

    1989-08-01

    Salt is the proposed host rock for geologic repositories of nuclear waste in several nations because it is nearly dry and probably impermeable. Although experiments and experience at potential salt sites indicate that salt may contain brine, the low porosity, creep, and permeability of salt make it still a good choice for geologic isolation. In this paper we summarize several mass-transfer and transport analyses of salt repositories. The mathematical details are given in our technical reports

  20. Technical support for GEIS: radioactive waste isolation in geologic formations. Volume 3. Stratigraphies of salt, granite, shale, and basalt

    International Nuclear Information System (INIS)

    1978-04-01

    This study presents the methodology and basic literature used to develop generic stratigraphic sections for the various geologic repository host rocks under considerations: salt, granite, shale and basalt

  1. An alternative hypothesis to the widely held view that renal excretion of sodium accounts for resistance to salt-induced hypertension

    Czech Academy of Sciences Publication Activity Database

    Kurtz, T. W.; DiCarlo, S. E.; Pravenec, Michal; Schmidlin, O.; Tanaka, M.; Morris Jr., R. C.

    2016-01-01

    Roč. 90, č. 5 (2016), s. 965-973 ISSN 0085-2538 Grant - others:AV ČR(CZ) AP1502 Program:Akademická prémie - Praemium Academiae Institutional support: RVO:67985823 Keywords : blood pressure * hypertension * kidney * salt * salt-resistance * salt-sensitivity * sodium * sodium chloride Subject RIV: FA - Cardiovascular Diseases incl. Cardiotharic Surgery Impact factor: 8.395, year: 2016

  2. Alternative methods to manage waste salt from repository excavation in the Deaf Smith County and Swisher County locations, Texas: A scoping study: Technical report

    International Nuclear Information System (INIS)

    1987-01-01

    This report describes and qualitatively evaluates eight options for managing the large volumes of salt and salt-laden rock that would result from the excavation of a high-level radioactive waste repository in Deaf Smith County or Swisher County, Texas. The options are: distribution for commercial use; ocean disposal; deep-well injection; disposal in multilevel mines on the site; disposal in abandoned salt mines off the site; disposal off the site in abandoned mines developed for minerals other than salt; disposal in excavated landfills; and surface disposal on alkali flats. The main features of each option are described, as well as the associated environmental and economic impacts, and regulatory constraints. The options are evaluated in terms of 11 factors that jointly constitute a test of relative suitability. The results of the evaluation and implications for further study are indicated. This document does not consider or include the actual numbers, findings, or conclusions contained in the final Deaf Smith County Environmental Assessment (DOE, 1986). 43 refs., 8 tabs

  3. Lithology, microstructures, fluid inclusions, and geochemistry of rock salt and of the cap-rock contact in Oakwood Dome, East Texas: significance for nuclear waste storage. Report of investigations No. 120

    International Nuclear Information System (INIS)

    Dix, O.R.; Jackson, M.P.A.

    1982-01-01

    Oakwood salt dome in Leon and Freestone Counties, Texas, has a core composed of a diapiric salt stock at a depth of 355 m. A vertical borehole in the center of the salt stock yielded 57.3 m of continuous rock-salt core overlain by 137 m of anhydrite-calcite cap rock. The lower 55.3 m of rock salt exhibits a strong, penetrative schistosity and parallel cleavage dipping at 30 to 40 0 and more than 60 variably dipping layers of disseminated anhydrite. Anhydrite constitutes 1.3 +- 0.7 percent of the rock-salt core. The upper 2 m of rock salt is unfoliated, comprising a lower 1.4-m interval of medium-grained granoblastic rock salt and an upper 0.6-m interval of coarse-grained granoblastic rock salt. An abrupt, cavity-free contact separates rock salt from laminated cap rock consisting of granoblastic-polygonal anhydrite virtually devoid of halite or pore space. Microstructures and concentration gradients of fluid inclusions suggest that the unfoliated rock salt at the crest of the salt stock was once strongly foliated, but that this fabric was destroyed by solid-state recrystallization. Downward movement of brine from the rock-salt - cap-rock contact was apparently accompanied by two recrystallization fronts. Dissolution of halite at the contact released disseminated anhydrite that presumably accumulated as sand on the floor of the dissolution cavity. Renewed rise of the salt stock closed the cavity, and the anhydrite sand was accreted against the base of the cap rock. Much, if not all, of the lamination in the 80 m of anhydrite cap rock may result from cycles of dissolution, recrystallization, and upward movement in the salt stock, followed by accretion of anhydrite residuum as laminae against the base of the cap rock. These processes, which are strongly influenced by fluids, act both to breach waste repositories and to geologically isolate them

  4. In situ-experiments on the disposal of high-level radioactive wastes (HAW) at the Asse salt mine Federal Republic of Germany

    International Nuclear Information System (INIS)

    Kuhn, K.; Rothfuchs, T.

    1989-01-01

    Deep geological salt formations are considered as being the most suitable medium for the disposal of radioactive wastes in the Federal Republic of Germany (FRG). This paper reports how, in order to develop and to prove the necessary disposal techniques, the Asse Salt Mine in the northern part of Germany is being used as a national R and D facility for the execution of representative in situ-tests. Besides the test-wise disposal of low-and medium-level radioactive waste, a series of in situ experiments was performed on the disposal of high-level radioactive waste (HAW). The so-called HAW repository is being performed from 1983 through 1994 will be the most important pilot test for the HAW repository in the FRG. During this experiment, 30 vitrified high-level radioactive heat and radiation sources will be emplaced in six underground boreholes. The duration of testing will be approximately five years. In addition to the investigations of the interactions of the heat and radiation sources and the host rock, a complete handling system for HAW-canisters is being developed and proved

  5. Thermoelastic/plastic analysis of waste-container sleeve: III. Influence of salt strength on sleeve loading. Technical memorandum report (RSI-0018)

    International Nuclear Information System (INIS)

    Pariseau, W.G.

    1975-01-01

    Three combinations of salt tensile, compressive and shear strength in linear and nonlinear yield conditions used in the axially symmetric, large displacement thermoelastic/plastic waste-container/sleeve loading estimates show no influence on the analysis. The salt remains elastic throughout the excavation and subsequent 10 year heating period. Tensile stresses are not observed, tensile strength is thus not important to the analysis even at 10 percent of the compressive strength value. Although strictly applicable only to the conditions of the analyses reported here, the capability for incorporating arbitrary strength combinations in linear or non-linear yield conditions is demonstrated. Computer plots of principal stresses and displacement fields at various stages of the excavation and heating simulation aid in the visualization of repository concept mechanics and show the possible need for additional mesh refinement for more precise stress information

  6. Renal transplantation in a patient with Bartter syndrome and glomerulosclerosis

    Science.gov (United States)

    Lee, Se Eun; Han, Kyoung Hee; Jung, Yun Hye; Lee, Hyun Kyung; Kang, Hee Gyung; Moon, Kyung Chul; Ha, Il Soo; Choi, Yong

    2011-01-01

    Bartter syndrome (BS) is a clinically and genetically heterogeneous inherited renal tube disorder characterized by renal salt wasting, hypokalemic metabolic alkalosis and normotensive hyperreninemic hyperaldosteronism. There have been several case reports of BS complicated by focal segmental glomerulosclerosis (FSGS). Here, we have reported the case of a BS patient who developed FSGS and subsequent end-stage renal disease (ESRD) and provided a brief literature review. The patient presented with classic BS at 3 months of age and developed proteinuria at 7 years. Renal biopsy performed at 11 years of age revealed a FSGS perihilar variant. Hemodialysis was initiated at 11 years of age, and kidney transplantation was performed at 16 years of age. The post-transplantation course has been uneventful for more than 3 years with complete disappearance of BS without the recurrence of FSGS. Genetic study revealed a homozygous p.Trp(TGG)610Stop(TGA) mutation in the CLCNKB gene. In summary, BS may be complicated by secondary FSGS due to the adaptive response to chronic salt-losing nephropathy, and FSGS may progress to ESRD in some patients. Renal transplantation in patients with BS and ESRD results in complete remission of BS. PMID:21359059

  7. Renal transplantation in a patient with Bartter syndrome and glomerulosclerosis

    Directory of Open Access Journals (Sweden)

    Se Eun Lee

    2011-01-01

    Full Text Available Bartter syndrome (BS is a clinically and genetically heterogeneous inherited renal tube disorder characterized by renal salt wasting, hypokalemic metabolic alkalosis and normotensive hyperreninemic hyperaldosteronism. There have been several case reports of BS complicated by focal segmental glomerulosclerosis (FSGS. Here, we have reported the case of a BS patient who developed FSGS and subsequent end-stage renal disease (ESRD and provided a brief literature review. The patient presented with classic BS at 3 months of age and developed proteinuria at 7 years. Renal biopsy performed at 11 years of age revealed a FSGS perihilar variant. Hemodialysis was initiated at 11 years of age, and kidney transplantation was performed at 16 years of age. The post-transplantation course has been uneventful for more than 3 years with complete disappearance of BS without the recurrence of FSGS. Genetic study revealed a homozygous p.Trp(TGG610Stop(TGA mutation in the CLCNKB gene. In summary, BS may be complicated by secondary FSGS due to the adaptive response to chronic salt-losing nephropathy, and FSGS may progress to ESRD in some patients. Renal transplantation in patients with BS and ESRD results in complete remission of BS.

  8. Salton Sea geothermal field as a natural analog for the near-field in a salt high-level nuclear waste repository

    International Nuclear Information System (INIS)

    Elders, W.A.; Moody, J.B.; Battelle Memorial Inst., Columbus, OH)

    1984-01-01

    The Salton Sea Geothermal Field (SSGF), on the delta of the Colorado River in southern California, is being studied as a natural analog for the near-field environment of proposed nuclear waste repositories in salt. A combination of mineralogical and geochemical methods is being employed to develop a three-dimenisonal picture of temperature, salinity, lithology, mineralogy, and chemistry of reactions between the reservoir rocks and the hot brines. Our aim is to obtain quantitative data on mineral stabilities and on mobilities of the naturally occurring radionuclides of concern in Commercial High-Level Waste (CHLW). These data will be used to validate the EQ3/6 geochemical code under development to model the salt near-field repository behavior. Maximum temperatures encountered in wells in the SSGF equal or exceed peak temperatures expected in a salt repository. Brines produced from these wells have major element chemistry similar to brines from candidate salt sites. Relative to the rocks, these brines are enriched in Na, Mn, Sr, Ra, and Po, depleted in Ba, Si, Mg, Ti, and Al, and strongly depleted in U and Th. However, the unaltered rocks contain only about 2 to 3 ppm of U and 4 to 12 ppm of Th, largely in detrital epidotes and zircons. Samples of hydrothermally altered rocks from a wide range of temperature and salinity show rather similar uniform low concentrations of these elements, even when authigenic illite, chlorite, ipidote and feldspar are present. These observations suggest that U and Th are relatively immobile in these hot brines. However, Ra, Po, Cs, and Sr are relatively mobile. Work is continuing to document naturally occurring radionuclide partitioning between SSGF minears and brine over a range of temperature, salinity, and lithology. 8 refs., 7 figs., 2 tabs

  9. Rheology Of MonoSodium Titanate (MST) And Modified Mst (mMST) Mixtures Relevant To The Salt Waste Processing Facility

    Energy Technology Data Exchange (ETDEWEB)

    Koopman, D. C.; Martino, C. J.; Shehee, T. C.; Poirier, M. R.

    2013-07-31

    The Savannah River National Laboratory performed measurements of the rheology of suspensions and settled layers of treated material applicable to the Savannah River Site Salt Waste Processing Facility. Suspended solids mixtures included monosodium titanate (MST) or modified MST (mMST) at various solid concentrations and soluble ion concentrations with and without the inclusion of kaolin clay or simulated sludge. Layers of settled solids were MST/sludge or mMST/sludge mixtures, either with or without sorbed strontium, over a range of initial solids concentrations, soluble ion concentrations, and settling times.

  10. Rheology Of MonoSodium Titanate (MST) And Modified Mst (mMST) Mixtures Relevant To The Salt Waste Processing Facility

    International Nuclear Information System (INIS)

    Koopman, D. C.; Martino, C. J.; Shehee, T. C.; Poirier, M. R.

    2013-01-01

    The Savannah River National Laboratory performed measurements of the rheology of suspensions and settled layers of treated material applicable to the Savannah River Site Salt Waste Processing Facility. Suspended solids mixtures included monosodium titanate (MST) or modified MST (mMST) at various solid concentrations and soluble ion concentrations with and without the inclusion of kaolin clay or simulated sludge. Layers of settled solids were MST/sludge or mMST/sludge mixtures, either with or without sorbed strontium, over a range of initial solids concentrations, soluble ion concentrations, and settling times

  11. Radiation induced F-center and colloid formation in synthetic NaCl and natural rock salt: applications to radioactive waste repositories

    International Nuclear Information System (INIS)

    Levy, P.W.; Loman, J.M.; Kierstead, J.A.

    1983-01-01

    Radiation damage, particularly Na metal colloid formation, has been studied in synthetic NaCl and natural rock salt using unique equipment for making optical absorption, luminescence and other measurements during irradiation with 1 to 3 MeV electrons. Previous studies have established the F-center and colloid growth phenomenology. At temperatures where colloids form most rapidly, 100 to 250 C, F-centers appear when the irradiation is initiated and increase at a decreasing rate to a plateau, reached at doses of 10 6 to 10 7 rad. Concomitant colloid growth is described by classical nucleation and growth curves with the transition to rapid growth occurring at 10 6 to 10 7 rad. The colloid growth rate is low at 100 C, increases markedly to a maximum at 150 to 175 C and decreases to a negligible rate at 225 C. At 1.2x10 8 rad/h the induction period is >10 4 sec at 100 C, 10 4 sec at 275 C. The colloid growth in salt from 14 localities is well described by C(dose)/sup n/ relations. Data on WIPP site salt (Los Medanos, NM, USA) has been used to estimate roughly the colloid expected in radioactive waste repositories. Doses of 1 to 2x10 10 rad, which will accumulate in salt adjacent to lightly shielded high level canisters in 200 to 500 years, will convert between 1 and 100% of the salt to Na colloids (and Cl) if back reactions or other limiting reactions do not occur. Each high level lightly shielded canister may ultimately be surrounded by 200 to 300 kg of colloid sodium. Low level or heavily shielded canisters may produce as little as 1 kg sodium

  12. Disposal of Radioactive Wastes in Natural Salt; Elimination des Dechets Radioactifs dans le Sel Naturel; 0423 0414 ; Evacuacion de Desechos Radiactivos en Formaciones Salinas Naturales

    Energy Technology Data Exchange (ETDEWEB)

    Parker, F. L.; Boegly, W. J.; Bradshaw, R. L.; Empson, F. M.; Hemphill, L.; Struxness, E. G.; Tamura, T. [Health Physics Division, Oak Ridge National Laboratory, Oak Ridge, TN (United States)

    1960-07-01

    The proposed use of cavities in salt formations as a disposal site for radioactive wastes is based upon : 1. Existence of salt for geologic time periods, 2. The impermeability of salt to the passage of water; 3. The widespread geographical distribution of salt; 4. The extremely large quantities of salt available; 5. The structural strength of salt; 6. The relatively high thermal conductivity of salt in comparison with other general geologic formations; 7. The possible recovery of valuable fission products in the wastes injected into the salt; 8. The relative ease of forming cavities in salt by mining, and the even greater ease and low cost of developing solution cavities in salt; and 9. The low seismicity in the areas of major salt deposits. Radioactive liquid wastes can be stored in cavities in natural salt formations if the structural properties of the salt are not adversely affected by chemical interaction, pressure, temperature, and radiation. Analytical studies show that it is possible to-store 2-year-old 10,000 MWD/T, 800 gal/ton waste in a sphere of 10 ft diameter without exceeding a temperature of 200 Degree-Sign F. Laboratory tests show that the structural properties and thermal conductivity of rock salt are not greatly altered by high radiation doses, although high temperatures increase the creep rate for both irradiated and unirradiated samples. Chemical interaction of liquid wastes with salt produces chlorine and other chlorine compound gases, but the volumes are not excessive. The migration of nuclides through the salt and deformation of the cavity and chamber can only be studied in undisturbed salt in situ. One-fifth-scale models have been run in a bedded salt deposit in Hutchinson, Kansas, and full-scale field tests are in progress. (author) [French] L'emploi envisage des cavites des gisements de sel comme lieu d'evacuation des dechets radioactifs se-fonde sur les considerations suivantes: 1. L'existence du sel dans des formations correspondant a

  13. Potential use of reverse osmosis in managing saltwater waste collected at road-salt storage facilites [sic].

    Science.gov (United States)

    2006-01-01

    The implementation of its anti-icing program comprises a large part of the Virginia Department of Transportation's (VDOT) maintenance effort. Earlier research confirmed that VDOT captures a large volume of salt-laden stormwater runoff at its 300+ sal...

  14. Generic aspects of salt repositories

    International Nuclear Information System (INIS)

    Laughon, R.B.

    1979-01-01

    The history of geological disposal of radioactive wastes in salt is presented from 1957 when a panel of the National Academy of Sciences-National Research Council recommended burial in bedded salt deposits. Early work began in the Kansas, portion of the Permian Basin where simulated wastes were placed in an abandoned salt mine at Lyons, Kansas, in the late 1960's. This project was terminated when the potential effect of nearby solution mining activities could not be resolved. Evaluation of bedded salts resumed a few years later in the Permian Basin in southeastern New Mexico, and search for suitable sites in the 1970's resulted in the formation of the National Waste Terminal Storage Program in 1976. Evaluation of salt deposits in many regions of the United States has been virtually completed and has shown that deposits having the greatest potential for radioactive waste disposal are those of the largest depositional basins and salt domes of the Gulf Coast region

  15. Radioactive waste isolation in salt: peer review of the D'Appolonia report on Schematic Designs for Penetration Seals for a Repository in the Permian Basin, Texas

    International Nuclear Information System (INIS)

    Hambley, D.F.; Stormont, J.C.; Russell, J.E.; Edgar, D.E.; Fenster, D.F.; Harrison, W.; Tisue, M.W.

    1984-09-01

    Argonne made the following recommedations for improving the reviewed reports. The authors of the report should: state the major assumptions of the study in Sec. 1.1 rather than later in the report; consider using salt for the shaft seals in salt horizons; reconsider whether keys are needed for the bulkheads; provide for interface grouting because use of expansive cement will not guarantee that interfaces will be impermeable; discuss the sealing schedule and, where appropriate, consider what needs to be done to ensure that emplaced radioactive waste could be retrieved if necessary; describe in more detail the sealing of the Dockum and Ogallala aquifers; consider an as low as reasonably achievable approach to performance requirements for the initial design phase; address the concerns in the 1983 US Nuclear Regulatory Commission document entitled Draft Technical Position: Borehole and Shaft Sealing of High-Level Nuclear Waste Repositories; cite the requirements for release of radioactivity by referring to specific clauses in the regulations of the US Environmental Protection Agency; and provide further explanation in the outline of future activities about materials development and verification testing. More emphasis on development of accelerated testing programs is also required

  16. Sodium Chloride Supplementation Is Not Routinely Performed in the Majority of German and Austrian Infants with Classic Salt-Wasting Congenital Adrenal Hyperplasia and Has No Effect on Linear Growth and Hydrocortisone or Fludrocortisone Dose.

    Science.gov (United States)

    Bonfig, Walter; Roehl, Friedhelm; Riedl, Stefan; Brämswig, Jürgen; Richter-Unruh, Annette; Fricke-Otto, Susanne; Hübner, Angela; Bettendorf, Markus; Schönau, Eckhard; Dörr, Helmut; Holl, Reinhard W; Mohnike, Klaus

    2018-01-01

    Sodium chloride supplementation in salt-wasting congenital adrenal hyperplasia (CAH) is generally recommended in infants, but its implementation in routine care is very heterogeneous. To evaluate oral sodium chloride supplementation, growth, and hydrocortisone and fludrocortisone dose in infants with salt-wasting CAH due to 21-hydroxylase in 311 infants from the AQUAPE CAH database. Of 358 patients with classic CAH born between 1999 and 2015, 311 patients had salt-wasting CAH (133 females, 178 males). Of these, 86 patients (27.7%) received oral sodium chloride supplementation in a mean dose of 0.9 ± 1.4 mmol/kg/day (excluding nutritional sodium content) during the first year of life. 225 patients (72.3%) were not treated with sodium chloride. The percentage of sodium chloride-supplemented patients rose from 15.2% in children born 1999-2004 to 37.5% in children born 2011-2015. Sodium chloride-supplemented and -unsupplemented infants did not significantly differ in hydrocortisone and fludrocortisone dose, target height-corrected height-SDS, and BMI-SDS during the first 2 years of life. In the AQUAPE CAH database, approximately one-third of infants with salt-wasting CAH receive sodium chloride supplementation. Sodium chloride supplementation is performed more frequently in recent years. However, salt supplementation had no influence on growth, daily fludrocortisone and hydrocortisone dose, and frequency of adrenal crisis. © 2017 S. Karger AG, Basel.

  17. Corrosion behaviour of selected high-level waste packaging materials under gamma irradiation and in-situ disposal conditions in rock salt

    International Nuclear Information System (INIS)

    Smailos, E.; Schwarzkopf, W.; Koester, R.

    1988-07-01

    Corrosion studies performed until now on a number of materials have shown that unalloyed steels, Hastelloy C4 and Ti 99.8-Pd are the most promising materials for a long-term resistant packaging to be used in high-level waste (HLW) canister disposal in rock salt formations. To characterize their corrosion behaviour in more detail, additional studies have been performed. The influence has been examined which is exerted by the gamma dose rate (1 Gy/h to 100 Gy/h) on the corrosion of three preselected steels and Hastelloy C4 at 90 0 C in a salt brine (Q-brine) rich in MgCl 2 , i.e., conditions relevant to accident scenarios in a repository. In addition, in-situ corrosion experiments have been carried out in the Asse salt mine at elevated temperatures (120 0 C to 210 0 C) in the absence and in the presence of a gamma radiation field of 3 x 10 2 Gy/h, within the framework of the German/US Brine Migration Test. Under the test conditions the gamma radiation did not exert a significant influence on the corrosion of the steels investigated, whereas Hastelloy C4, exposed to dose rates of 10 Gy/h and 100 Gy/h, underwent pitting and crevice corrosion (20 μm/a at the maximum).The low amounts of migrated salt brine (140 ml after 900 days) in the in-situ- experiment did not produce noticeable corrosion of the materials. (orig./RB) [de

  18. Numerical and experimental investigations on the time dependent behavior of a salt dome with a high-level waste repository

    International Nuclear Information System (INIS)

    Prij, J.; Vons, L.H.

    1984-01-01

    Results are presented of in-situ measurements, performed in a 300 m deep dry-drilled borehole, in the ASSE-mine. Convergence measurements at ambient as well as elevated temperatures and pressure measurements at elevated temperatures are discussed. Creep equations derived from these experiments are used for the numerical analysis of the time dependent behavior of a salt dome with a HLW repository. The analyses show that the total stresses in the salt remain compressive with deviatoric components smaller than 3 MPa. 9 references, 6 figures, 1 table

  19. Radioactive waste isolation in salt: Peer review of the Office of Nuclear Waste Isolation's draft report on a multifactor test design to investigate uniform corrosion of low-carbon steel

    Energy Technology Data Exchange (ETDEWEB)

    Paddock, R.A.; Lerman, A.; Ditmars, J.D.; Macdonald, D.D.; Peerenboom, J.P.; Was, G.S.; Harrison, W.

    1987-01-01

    This report documents Argonne National Laboratory's review of an internal technical memorandum prepared by Battelle Memorial Institute's Office of Nuclear Waste Isolation (ONWI) entitled Multifactor Test Design to Investigate Uniform Corrosion of Low-Carbon Steel in a Nuclear Waste Salt Repository Environment. The several major areas of concern identified by peer review panelists are important to the credibility of the test design proposed in the memorandum and are to adequately addressed there. These areas of concern, along with specific recommendations to improve their treatment, are discussed in detail in Sec. 2 of this report. The twenty recommendations, which were abstracted from those discussions, are presented essentially in the order in which they are introduced in Sec. 2.

  20. Radioactive waste isolation in salt: Peer review of the Office of Nuclear Waste Isolation's draft report on a multifactor test design to investigate uniform corrosion of low-carbon steel

    International Nuclear Information System (INIS)

    Paddock, R.A.; Lerman, A.; Ditmars, J.D.; Macdonald, D.D.; Peerenboom, J.P.; Was, G.S.; Harrison, W.

    1987-01-01

    This report documents Argonne National Laboratory's review of an internal technical memorandum prepared by Battelle Memorial Institute's Office of Nuclear Waste Isolation (ONWI) entitled Multifactor Test Design to Investigate Uniform Corrosion of Low-Carbon Steel in a Nuclear Waste Salt Repository Environment. The several major areas of concern identified by peer review panelists are important to the credibility of the test design proposed in the memorandum and are to adequately addressed there. These areas of concern, along with specific recommendations to improve their treatment, are discussed in detail in Sec. 2 of this report. The twenty recommendations, which were abstracted from those discussions, are presented essentially in the order in which they are introduced in Sec. 2

  1. National waste terminal storage repository in a bedded salt formation for spent unreprocessed fuel. Special study No. 3. Waste retrieval from backfilled regions

    International Nuclear Information System (INIS)

    1978-09-01

    Methods and costs were studied for delayed canister retrieval from rooms that had been backfilled immediately after canister storage. The effects of this method of storage on mine geometry, thermal and rock mechanics environments, mine development and operations, mine ventilation, time schedule, retrieval machinery and safety were investigated. Salt and air temperatures were determined. Pillar width, number of rooms, extraction ratio, tonnages of mined salt, and salt handling and hoisting requirements were calculated. The required changes in mining equipment were established. Salt handling and elapsed time schedules were developed. Ventilation requirements - size and number of shafts, size the arrangement of airways, number of stacks, and size and number of fans were then calculated. The development sequence of these facilities was established. Canister retrieval problems were analyzed for canisters stuck in the hole as well as free. Retrieval methods and machinery were studied and are described. Safety with respect to both radiation and room collapse was studied and compared with CDR safety conditions. The effects of a reduced themal loading of 30 KW/acre on temperatures, room closure, mine layout, ventilation and ground control were studied and reported. A cost estimate was prepared, giving cost differentials between the base CDR costs and Special Study No. 3. Two appendices are included. The first contains nine Heat Transfer memoranda that state the thermal basis of this study. The second appendix provides a detailed operating time analysis of the retrieval machinery

  2. [Hypertension and renal disease

    DEFF Research Database (Denmark)

    Kamper, A.L.; Pedersen, E.B.; Strandgaard, S.

    2009-01-01

    Renal mechanisms, in particular the renin-angiotensin system and renal salt handling, are of major importance in blood pressure regulation. Co-existence of hypertension and decreased renal function may be due to nephrosclerosis secondary to hypertension, or primary renal disease with secondary...... hypertension. Mild degrees of chronic kidney disease (CKD) can be detected in around 10% of the population, and detection is important as CKD is an important risk factor for atherosclerotic cardiovascular disease. Conversely, heart failure may cause an impairment of renal function. In chronic progressive...... nephropathy, effective blood pressure lowering is of paramount importance, and angiotensin converting enzyme inhibitors and angiotensin receptor blockers are agents of choice Udgivelsesdato: 2009/6/15...

  3. Long-term interactions of full-scale cemented waste simulates in salt solutions. Summary report; Langzeit-Wechselwirkungen von zementierten Abfallsimulaten im Originalmassstab mit Salzloesungen. Zusammenfassender Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Kienzler, Bernhard; Borkel, Christoph; Metz, Volker [Karlsruher Institut fuer Technologie (KIT), Karlsruhe (Germany). Inst. fuer Nukleare Entsorgung (INE)

    2015-12-04

    On March 13,.2013 the Federal Office of Radiation Protection (BfS) published a note that the responsible group of the Helmholtz Zentrums Muenchen had finished the experiments in the socalled leaching test room at the 490 m level of the Asse II mine. In this room, the previous operator the Gesellschaft fuer Strahlen- und Umweltforschung mbH (GSF) carried out leaching and corrosion experiments with cemented full-scale samples. These experiments were performed since 1979 requested by the licensing authorities. With respect to the safety case for the Asse salt mine it was a need to demonstrate the transferability of results obtained by laboratory samples to real waste forms and to investigate the effects of the industrial cementation process an the properties of the waste forms. A research program was initiated by the Nuclear Research Centre Karlsruhe (today Karlsruhe Institute of Technology, KIT) and the Institut fuer Tieflagerung of the Gesellschaft fuer Strahlenforschung m.b.H. (GSF). Since 1996 the scientific supervision of the experiments were dedicated to the Institute for Nuclear Waste Disposal (INE) of KIT. Until 2013, the corroding solutions were sampled several times. In 2006 four full-scale samples were retrieved and investigated with respect to variations of the solids. After termination of the experiments in January 2013, radioactively doped samples were transferred to KIT-INE for final evaluations. The present report summarizes the background and objectives of the experiments as well as the results of the solutions and solid state analyses.

  4. Disposal of high-level waste from nuclear power plants in Denmark. Salt dome investigations. v.1

    International Nuclear Information System (INIS)

    1981-01-01

    A summary is presented of a report in five volumes on possible disposal of radioactive waste in Denmark. The investigation was made by the Danish electric utilities ELKRAFT and ELSAM at the request of the Danish Government. The investigation proved it possible to consider two alternative designs for a disposal facility, one based on the deposition of waste in individual, deep holes, the other on placing the waste in mine galleries. A safety analysis was completed with the Mors dome as example. The purpose of the analysis was to prove whether safe disposal of high-level waste in Denmark was feasible. The utilities concluded that the results of the analysis were satisfactory and the report is now being assessed by the authorities. (BP)

  5. De-chlorination and solidification of radioactive LiCl waste salt by using SiO_2-Al_2O_3-P_2O_5 (SAP) inorganic composite including B_2O_3 component

    International Nuclear Information System (INIS)

    Lee, Ki Rak; Park, Hwan-Seo; Cho, In-Hak; Choi, Jung-Hoon; Eun, Hee-Chul; Lee, Tae-Kyo; Han, Seung Youb; Ahn, Do-Hee

    2017-01-01

    SAP (SiO_2-Al_2O_3-P_2O_5) composite has been recently studied in KAERI to deal with the immobilization of radioactive salt waste, one of the most problematic wastes in the pyro-chemical process. Highly unstable salt waste was successfully converted into stable compounds by the dechlorination process with SAPs, and then a durable waste form with a high waste loading was produced when adding glassy materials to dechlorination product. In the present study, U-SAP composite which is SAP bearing glassy component (Boron) was synthesized to remove the adding and mixing steps of glassy materials for a monolithic wasteform. With U-SAPs prepared by a sol-gel process, a series of wasteforms were fabricated to identify a proper reaction condition. Physical and chemical properties of dechlorination products and U-SAP wasteforms were characterized by XRD, DSC, SEM, TGA and PCT-A. A U-SAP wasteform showed suitable properties as a radioactive wasteform such as dense surface morphology, high waste loading, and high durability at the optimized U-SAP/salt ratio 2.

  6. Thermoelastic analysis of spent fuel and high level radioactive waste repositories in salt. A semi-analytical solution

    International Nuclear Information System (INIS)

    St John, C.M.

    1977-04-01

    An underground repository containing heat generating, High Level Waste or Spent Unreprocessed Fuel may be approximated as a finite number of heat sources distributed across the plane of the repository. The resulting temperature, displacement and stress changes may be calculated using analytical solutions, providing linear thermoelasticity is assumed. This report documents a computer program based on this approach and gives results that form the basis for a comparison between the effects of disposing of High Level Waste and Spent Unreprocessed Fuel

  7. Stabilization/Solidification of Radioactive LiCl-KCl Waste Salt by Using SiO2-Al2O3-P2O5(SAP) Inorganic Composite: Part 2. The Effect of SAP Composition on Stabilization/Solidification

    International Nuclear Information System (INIS)

    Ahn, Soo Na; Park, Hwan Seo; Cho, In Hak; Kim, In Tae; Cho, Yong Zun

    2012-01-01

    Metal chloride waste is generated as a main waste streams in a series of electrolytic processes of a pyrochemical process. Different from carbonate or nitrate salt, metal chloride is not decomposed into oxide and chlorine but it is just vaporized. Also, it has low compatibility with conventional silicate glasses. Our research group adapted the dechlorination approach for the immobilization of waste salt. In this study, the composition of SAP (SiO 2 -Al 2 O 3 -P 2 O 5 ) was adjusted to enhance the reactivity and to simplify the solidification process as a subsequent research. The addition of Fe 2 O 3 into the basic SAP decreased the SAP/Salt ratio in weight from 3 for SAP 1071 to 2.25 for M-SAP(Fe=0.1). The experimental results indicated that the addition of Fe 2 O 3 increased the reactivity of M-SAP with LiCl-KCl but the reactivity gradually decreased above Fe=0.1. Also, introducing B 2 O 3 into M-SAP requires no glass binder for the consolidation of reaction products. U-SAP (SiO 2 -Al 2 O 3 -P 2 O 5 ) could effectively dechlorinate the LiCl-KCl waste and its reaction product could be consolidated as a monolithic form without a glass binder. The leaching test result indicated that U-SAP 1071 was more durable than other SAPs wasteform. By using U-SAP, 1 g of waste salt could generated 3 - 4 g of wasteform for final disposal. The final volume would be about 3 - 4 times lower than the glass-bonded sodalite. From these results, it could be concluded that the dechlorination approach using U-SAP would be one of prospective methods to manage the volatile waste salt.

  8. Review of applicable technology: solution mining of caverns in salt domes to serve as repositories for radioactive wastes

    International Nuclear Information System (INIS)

    1976-01-01

    There is an abundance of salt domes in the Gulf Coastal region. Advances in leaching technology and cavern shape control make it possible to build large caverns with configurations approaching teardrops, cylinders, and spheres. Fenix and Scisson has designed and constructed several dozen caverns in sizes up to three million barrels (16.8 million cubic feet). It is now within current technological bounds to evacuate the brine left in the cavern following construction, dehumidify the cavern atmosphere and supply conditioned cavern ventilation. The state-of-the-art in drilling large diameter holes has advanced to the point that it is now possible to drill 120-in. holes as deep as 6,000 ft and 144-in. holes to lesser depths. Additional research is needed in the area of cavern stability. Cavern shrinkage rates are known to increase with depth because of lower salt strengths at higher pressures and temperatures

  9. Investigation of the utility of Gulf Coast salt domes for the storage or disposal of radioactive wastes

    International Nuclear Information System (INIS)

    Martinez, J.D.; Thoms, R.L.; Kupfer, D.H.

    1976-01-01

    Analysis of tectonic, geohydrologic, and cultural data led to the selection of three salt domes (Vacherie, Rayburn's, Prothro) in the North Louisiana Basin and three (Keechi, Mt. Sylvan, Palestine) in the Northeast Texas Basin. Results of the tectonic stability studies (monitoring of dome movement, quaternary, Mesozoic and Tertiary, seismic, corehole in Vacherie) and hydrologic stability studies (numerical modeling, caprock, mine leaks) are discussed in detail and recommendations for further study are given

  10. Investigation of the utility of Gulf Coast salt domes for the storage or disposal of radioactive wastes

    Energy Technology Data Exchange (ETDEWEB)

    Martinez, J.D.; Thoms, R.L.; Kupfer, D.H.

    1976-09-30

    Analysis of tectonic, geohydrologic, and cultural data led to the selection of three salt domes (Vacherie, Rayburn's, Prothro) in the North Louisiana Basin and three (Keechi, Mt. Sylvan, Palestine) in the Northeast Texas Basin. Results of the tectonic stability studies (monitoring of dome movement, quaternary, Mesozoic and Tertiary, seismic, corehole in Vacherie) and hydrologic stability studies (numerical modeling, caprock, mine leaks) are discussed in detail and recommendations for further study are given. (DLC)

  11. Status and future developments of risk analysis for repositories of radioactive wastes in salt formations in the Federal Republic of Germany

    International Nuclear Information System (INIS)

    Proske, R.

    1977-01-01

    In the Federal Republic of Germany a hypothetical repository for High-Level Radioactive Waste in a salt formation was taken as basis for a first attempt to use the methodology of risk analysis in order to get statements on the safety of such a geologic and mining system. Several institutions were engaged in drawing up fault trees, development of release models and calculation of the risk. A lot of R+D-work is scheduled to be carried out in future which includes optimization of the application of risk analysis methodology to geologic and mining systems, further development of release models, development of a model describing the migration of radionuclides in typical geologic strata and soils of the Federal Republic of Germany and application of risk analysis methodology to different repositories and disposal technologies

  12. The Microbiology of Subsurface, Salt-Based Nuclear Waste Repositories: Using Microbial Ecology, Bioenergetics, and Projected Conditions to Help Predict Microbial Effects on Repository Performance

    International Nuclear Information System (INIS)

    Swanson, Juliet S.; Cherkouk, Andrea; Arnold, Thuro; Meleshyn, Artur; Reed, Donald T.

    2016-01-01

    This report summarizes the potential role of microorganisms in salt-based nuclear waste repositories using available information on the microbial ecology of hypersaline environments, the bioenergetics of survival under high ionic strength conditions, and ''repository microbiology'' related studies. In areas where microbial activity is in question, there may be a need to shift the research focus toward feasibility studies rather than studies that generate actual input for performance assessments. In areas where activity is not necessary to affect performance (e.g., biocolloid transport), repository-relevant data should be generated. Both approaches will lend a realistic perspective to a safety case/performance scenario that will most likely underscore the conservative value of that case.

  13. The Microbiology of Subsurface, Salt-Based Nuclear Waste Repositories: Using Microbial Ecology, Bioenergetics, and Projected Conditions to Help Predict Microbial Effects on Repository Performance

    Energy Technology Data Exchange (ETDEWEB)

    Swanson, Juliet S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Cherkouk, Andrea [Helmholtz-Zentrum Dresden-Rossendorf, Rossendorf (Germany); Arnold, Thuro [Helmholtz-Zentrum Dresden-Rossendorf, Rossendorf (Germany); Meleshyn, Artur [Gesellschaft fur Anlagen und Reaktorsicherheit, Braunschweig (Germany); Reed, Donald T. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-11-17

    This report summarizes the potential role of microorganisms in salt-based nuclear waste repositories using available information on the microbial ecology of hypersaline environments, the bioenergetics of survival under high ionic strength conditions, and “repository microbiology” related studies. In areas where microbial activity is in question, there may be a need to shift the research focus toward feasibility studies rather than studies that generate actual input for performance assessments. In areas where activity is not necessary to affect performance (e.g., biocolloid transport), repository-relevant data should be generated. Both approaches will lend a realistic perspective to a safety case/performance scenario that will most likely underscore the conservative value of that case.

  14. Summary of computational support and general documentation for computer code (GENTREE) used in Office of Nuclear Waste Isolation Pilot Salt Site Selection Project

    International Nuclear Information System (INIS)

    Beatty, J.A.; Younker, J.L.; Rousseau, W.F.; Elayat, H.A.

    1983-01-01

    A Decision Tree Computer Model was adapted for the purposes of a Pilot Salt Site Selection Project conducted by the Office of Nuclear Waste Isolation (ONWI). A deterministic computer model was developed to structure the site selection problem with submodels reflecting the five major outcome categories (Cost, Safety, Delay, Environment, Community Impact) to be evaluated in the decision process. Time-saving modifications were made in the tree code as part of the effort. In addition, format changes allowed retention of information items which are valuable in directing future research and in isolation of key variabilities in the Site Selection Decision Model. The deterministic code was linked to the modified tree code and the entire program was transferred to the ONWI-VAX computer for future use by the ONWI project

  15. Heavy metals, salts and organic residues in solid urban waste landfills and surface waters in their discharge areas: determinants for restoring their discharge areas: determinants for restoring their impact

    International Nuclear Information System (INIS)

    Hernandez, A. J.; Pastor, J.

    2009-01-01

    This report describes a continuous assessment of the impact of solid urban waste (SUW) landfills in the central Iberian Peninsula that were sealed with a layer of soil 20 years ago. cover soils and soils from discharge areas have been periodically analysed. Soil concentrations of salts and heavy metals affect the biotic components of these ecosystems. (Author)

  16. Studies of mechanisms and processes of relevance to the safety of nuclear waste repositories, as carried out prior to, during and after flovelling of the Hope potash salt mine

    International Nuclear Information System (INIS)

    1985-01-01

    Studies on the effects of a hypothetical accident involving water or brine intrusion into a waste repository in a salt mine are of special importance within the framework of safety assessments of salt formations as candidate sites for nuclear waste repositories. The measuring activities under review include the following: Physicochemical measurements for determining dissolution and recipitation of salts, transport mechanisms, temperature curves, natural build-up and efficiency of geochemical barriers in the brine. Geochemical measurements for obtaining information on the rock deformation prior to, during, and after flovelling. Geophysical measurements of microseismic behaviour of rock masses prior to, during, and after flovelling. Examination of an artificial barrier structure for the testing and assessment of technical barriers and their efficiency. (orig./HP) [de

  17. A natural analogue for near-field behaviour in a high level radioactive waste repository in salt: the Salton Sea geothermal field, California, USA

    International Nuclear Information System (INIS)

    Elders, W.A.

    1987-01-01

    In the Salton Sea Geothermal Field (SSGF), in the sediments of the delta of the Colorado River, we are developing a three-dimensional picture of active water/rock reactions at temperatures of 0 C and salinities of 7 to 25 weight percent to produce quantitative data on mineral stabilities and mobilities of naturally-occurring radio-nuclides. The aim is to produce data to validate geochemical computer codes being developed to assess the performance of a Commercial High-Level Waste (CHLW) repository in salt. Among the findings to date are: (1) greenschist facies metamorphism is occurring; (2) brine compositions are fairly similar to those expected in candidate salt repository sites; (3) U and Th concentrations in the rocks are typical for sedimentary rocks; (4) the brines are enriched in Na, Mn, Zn, Sr, Ra Po and strongly depleted in U and Th relative to the rocks; (5) significant radioactive disequilibria exist in brines and solid phases of the SSGF. The disequilibria in the actinide series allow estimation of the rates of brine-rock interaction and understanding of hydrologic processes and radionuclide behaviour. Work is continuing emphasizing the reactions of authigenic clay minerals, epidotes, feldspars, chlorites and sulphates. So far, adapting geochemical codes to the necessary combination of high salinity and high temperature has lagged behind the natural analogue study of the SSGF so that validation is still in progress. In the future our data can be also used in validating performance assessment codes which couple geochemistry and transport processes, and in design of waste packages and back fill compositions. (author)

  18. Location-independent study concerning the construction, operation and closure of possible facilities for the final storage of radioac