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Sample records for wasteforms

  1. Radionuclide Retention in Concrete Wasteforms

    Energy Technology Data Exchange (ETDEWEB)

    Wellman, Dawn M.; Jansik, Danielle P.; Golovich, Elizabeth C.; Cordova, Elsa A.

    2012-09-24

    Assessing long-term performance of Category 3 waste cement grouts for radionuclide encasement requires knowledge of the radionuclide-cement interactions and mechanisms of retention (i.e., sorption or precipitation); the mechanism of contaminant release; the significance of contaminant release pathways; how wasteform performance is affected by the full range of environmental conditions within the disposal facility; the process of wasteform aging under conditions that are representative of processes occurring in response to changing environmental conditions within the disposal facility; the effect of wasteform aging on chemical, physical, and radiological properties; and the associated impact on contaminant release. This knowledge will enable accurate prediction of radionuclide fate when the wasteforms come in contact with groundwater. Data collected throughout the course of this work will be used to quantify the efficacy of concrete wasteforms, similar to those used in the disposal of LLW and MLLW, for the immobilization of key radionuclides (i.e., uranium, technetium, and iodine). Data collected will also be used to quantify the physical and chemical properties of the concrete affecting radionuclide retention.

  2. CURRENT ANSTO RESEARCH ON WASTEFORM DEVELOPMENT

    Energy Technology Data Exchange (ETDEWEB)

    Vance, E.R.; Perera, D.S.; Stewart, M.W.A.; Begg, B.D.; Carter, M.L.; Day, R.A.; Moricca, S.; Smith, K.L.; Lumpkin, G.R.; Hanna, J.V.

    2003-02-27

    Current ANSTO scientific research on wasteform development for mainly high-level radioactive waste is directed towards practical applications. Titanate wasteform products we have developed or are developing are aimed at immobilization of: (a) tank wastes and sludges; (b) U-rich wastes from radioisotope production from reactor irradiation of UO2 targets; (c) Al-rich wastes arising from reprocessing of Al-clad fuels; (d) 99Tc; (e) high- Mo wastes arising from reprocessing of U-Mo fuels and (f) partitioned Cs-rich wastes. Other wasteforms include encapsulated zeolites or silica/alumina beads for immobilization of 129I. Wasteform production techniques cover hot isostatic and uniaxial pressing, sintering, and cold-crucible melting. In addition, building on previous work on speciation and leach resistance of Cs in cementitious products, we are studying geopolymers. Although we have a strong focus on candidate wasteforms for actual wastes, we have a considerable program directed at basic understanding of the wasteforms in regard to crystal chemistry, their dissolution behavior in aqueous media, radiation damage effects and processing techniques.

  3. Radionuclide Retention in Concrete Wasteforms - FY13

    Energy Technology Data Exchange (ETDEWEB)

    Snyder, Michelle MV; Golovich, Elizabeth C.; Wellman, Dawn M.; Crum, Jarrod V.; Lapierre, Robert; Dage, Denomy C.; Parker, Kent E.; Cordova, Elsa A.

    2013-10-15

    Assessing long-term performance of Category 3 waste cement grouts for radionuclide encasement requires knowledge of the radionuclide-cement interactions and mechanisms of retention (i.e., sorption or precipitation); the mechanism of contaminant release; the significance of contaminant release pathways; how wasteform performance is affected by the full range of environmental conditions within the disposal facility; the process of wasteform aging under conditions that are representative of processes occurring in response to changing environmental conditions within the disposal facility; the effect of wasteform aging on chemical, physical, and radiological properties; and the associated impact on contaminant release. This knowledge will enable accurate prediction of radionuclide fate when the wasteforms come in contact with groundwater. Data collected throughout the course of this work will be used to quantify the efficacy of concrete wasteforms, similar to those used in the disposal of low-level waste and mixed low-level waste, for the immobilization of key radionuclides (i.e., uranium, technetium, and iodine). Data collected will also be used to quantify the physical and chemical properties of the concrete affecting radionuclide retention.

  4. Nuclear wasteform materials: Atomistic simulation case studies

    Energy Technology Data Exchange (ETDEWEB)

    Chroneos, A., E-mail: alex.chroneos@open.ac.uk [Materials Engineering, The Open University, Milton Keynes MK7 6AA (United Kingdom); Department of Materials, Imperial College London, London SW7 2AZ (United Kingdom); Institute of Materials Science, NCSR Demokritos, GR-15310 Athens (Greece); Rushton, M.J.D. [Department of Materials, Imperial College London, London SW7 2AZ (United Kingdom); Jiang, C. [State Key Laboratory of Powder Metallurgy, Central South University, Changsha 410083 (China); Tsoukalas, L.H. [Department of Nuclear Engineering, Purdue University, West Lafayette, IN 47907 (United States)

    2013-10-15

    Ever increasing global energy demand combined with a requirement to reduce CO{sub 2} emissions has rekindled an interest in nuclear power generation. In order that nuclear energy remains publicly acceptable and therefore a sustainable source of power it is important that nuclear waste is dealt with in a responsible manner. To achieve this, improved materials for the long-term immobilisation of waste should be developed. The extreme conditions experienced by nuclear wasteforms necessitate the detailed understanding of their properties and the mechanisms acting within them at the atomic scale. This latter issue is the focus of the present review. Atomic scale simulation techniques can accelerate the development of new materials for nuclear wasteform applications and provide detailed information on their physical properties that cannot be easily accessed by experiment. The present article introduces examples of how atomic scale, computational modelling techniques have led to an improved understanding of current nuclear wasteform materials and also suggest how they may be used in the development of new wasteforms.

  5. Phase evolution in zirconolite glass-ceramic wasteforms

    Science.gov (United States)

    Maddrell, Ewan R.; Paterson, Hannah C.; May, Sarah E.; Burns, Kerry M.

    2017-09-01

    The evolution of crystalline phases in a model glass-ceramic wasteform system has been studied as a function of temperature and time. The work has shown that perovskite and sphene form as transient phases before final formation of zirconolite. The study also suggests some evidence for subtle structural transformations within the zirconolite phase.

  6. Characteristics of wasteform composing of phosphate and silicate to immobilize radioactive waste salts.

    Science.gov (United States)

    Park, Hwan-Seo; Cho, In-Hak; Eun, Hee Chul; Kim, In-Tae; Cho, Yong Zun; Lee, Han-Soo

    2011-03-01

    In the radioactive waste management, metal chloride wastes from a pyrochemical process is one of problematic wastes not directly applicable to a conventional solidification process. Different from a use of minerals or a specific phosphate glass for immobilizing radioactive waste salts, our research group applied an inorganic composite, SAP (SiO(2)-Al(2)O(3)-P(2)O(5)), to stabilize them by dechlorination. From this method, a unique wasteform composing of phosphate and silicate could be fabricated. This study described the characteristic of the wasteform on the morphology, chemical durability, and some physical properties. The wasteform has a unique "domain-matrix" structure which would be attributed to the incompatibility between silicate and phosphate glass. At higher amounts of chemical binder, "P-rich phase encapsulated by Si-rich phase" was a dominant morphology, but it was changed to be Si-rich phase encapsulated by P-rich phase at a lower amount of binder. The domain and subdomain size in the wasteform was about 0.5-2 μm and hundreds of nm, respectively. The chemical durability of wasteform was confirmed by various leaching test methods (PCT-A, ISO dynamic leaching test, and MCC-1). From the leaching tests, it was found that the P-rich phase had ten times lower leach-resistance than the Si-rich phase. The leach rates of Cs and Sr in the wasteform were about 10(-3)g/m(2)· day, and the leached fractions of them were about 0.04% and 0.06% at 357 days, respectively. Using this method, we could stabilize and solidify the waste salt to form a monolithic wasteform with good leach-resistance. Also, the decrease of waste volume by the dechlorination approach would be beneficial in the final disposal cost, compared with the present immobilization methods for waste salt.

  7. Characterization of monazite glass-ceramics as wasteform for simulated {alpha}-HLLW

    Energy Technology Data Exchange (ETDEWEB)

    He Yong [Materials Science and Chemical Engineering College, China University of Geosciences, Wuhan 430074 (China)], E-mail: heyongyu@263.net; Lue Yanjie; Zhang Qian [Materials Science and Chemical Engineering College, China University of Geosciences, Wuhan 430074 (China)

    2008-05-31

    Two monazite glass-ceramic wasteforms were sintered by mixing the lanthanum metaphosphate glass powder with the oxide powder of the components in simulated {alpha}-HLWs. The co-existence of components Al and Mo in an iron phosphate melt separated the melt into two immiscible glass melts, namely aluminum iron phosphate glass (Gb) and molybdenum iron phosphate glass (Gg). 24 wt% of ZrO{sub 2}, together with P{sub 2}O{sub 5} and proper amounts of Fe and Mo formed a zirconium pyrophosphate glass (Gg1), which was immiscible with the phase Gg. The iron ions in the wasteforms were all in Fe{sup 3+}, 1/3 of which was in 4-fold coordination. The O/P and O/(P + 1/3Fe{sup 3+}) ratios for the glass phases were Gg1 3.70, Gb 3.89-3.98, Gg 4.23-4.25, and Gg1 3.58, Gb 3.47-3.42, Gg 3.74-3.69, respectively. The dissolution rates of two wasteforms were 0.3008 and 0.2598 g/m{sup 2}d, respectively.

  8. Effects of gamma-ray irradiation on leaching of simulated 133Cs+ radionuclides from geopolymer wasteforms

    Science.gov (United States)

    Deng, Ning; An, Hao; Cui, Hao; Pan, Yang; Wang, Bing; Mao, Linqiang; Zhai, Jianping

    2015-04-01

    Leaching of simulated 133Cs+ radionuclides from geopolymer wasteforms was examined with regard to effects from gamma-ray irradiation. Specifically, the compressive strengths, microstructures, pore structures, and leaching resistance of geopolymer wasteforms before and after irradiation were characterized. The leaching experiments were performed by immersion of wasteforms in deionized water, ground water, and seawater. It was found that gamma rays did not produce significant morphological changes, except for changes in the pore size distribution. The cumulative leaching fraction of all the leachants from the irradiated samples increased relative to the non-radiated samples, particularly during long leaching periods (11-42 days). These results, and those from a mercury intrusion porosimeter analysis, can be attributed to irradiation-induced changes in pore structure. All the leaching indexes were greater than the minimum acceptable value of 6.0 set by the American Nuclear Society Standards committee, which indicated that the fly-ash geopolymers are suitable for radionuclide immobilization. However, the effects of gamma-ray irradiation on the immobilization of radionuclides cannot be ignored.

  9. Characterisation of Al corrosion and its impact on the mechanical performance of composite cement wasteforms by the acoustic emission technique

    Science.gov (United States)

    Spasova, L. M.; Ojovan, M. I.

    2008-04-01

    In this study acoustic emission (AE) non-destructive method was used to evaluate the mechanical performance of cementitious wasteforms with encapsulated Al waste. AE waves generated as a result of Al corrosion in small-size blast furnace slag/ordinary Portland cement wasteforms were recorded and analysed. The basic principles of the conventional parameter-based AE approach and signal-based analysis were combined to establish a relationship between recorded AE signals and different interactions between the Al and the encapsulating cement matrix. The AE technique was shown as a potential and valuable tool for a new area of application related to monitoring and inspection of the mechanical stability of cementitious wasteforms with encapsulated metallic wastes such as Al.

  10. Microstructure and leaching durability of glass composite wasteforms for spent clinoptilolite immobilisation

    Science.gov (United States)

    Juoi, J. M.; Ojovan, M. I.; Lee, W. E.

    2008-01-01

    Simulated spent Cs-clinoptilolite waste was immobilised in a monolithic glass composite material (GCM) produced by a pressureless sintering at 750 °C for 2 h duration. The effects of waste loading from 1:1 up to 1:10 glass to waste volume ratio (37 up to 88 wt%) on the GCM wasteform microstructure and leaching properties were analysed. The open porosity ranged between 0.84 and ˜13.2 % for the highest waste load. Significant changes occurred in the microstructure, phases present and wasteform durability at different waste loading. At waste loading up to 73 wt% of spent clinoptilolite, the GCM microstructure consists of several crystalline phases (clinoptilolite, sodalite, wollastonite and CsCl) that were fully encapsulated by a glass matrix. This leads to a low normalized leaching rate of Cs (remaining below 6.35 × 10 -6 g/cm 2 day in a GCM with 73 wt% waste) during a leaching test for 7 days conducted using ASTM C1220-98. In GCM's with waste loading exceeding 73 wt%, the crystalline phases present (clinoptilolite and CsCl) were not fully encapsulated by the glass matrix hence the normalized leaching rate of Cs was as high as 9.06 × 10 -4 g/cm 2 day at waste loading of ⩾80 wt%.

  11. Microstructure and leaching durability of glass composite wasteforms for spent clinoptilolite immobilisation

    Energy Technology Data Exchange (ETDEWEB)

    Juoi, J.M. [Immobilisation Science Laboratory, University of Sheffield, Sir Robert Hadfield Building, Mappin Street, Sheffield S1 3JD (United Kingdom)], E-mail: j.juoi@sheffield.ac.uk; Ojovan, M.I. [Immobilisation Science Laboratory, University of Sheffield, Sir Robert Hadfield Building, Mappin Street, Sheffield S1 3JD (United Kingdom); Lee, W.E. [Department of Materials, Imperial College London, S Kensington Campus, London SW7 2AZ (United Kingdom)

    2008-01-31

    Simulated spent Cs-clinoptilolite waste was immobilised in a monolithic glass composite material (GCM) produced by a pressureless sintering at 750 deg. C for 2 h duration. The effects of waste loading from 1:1 up to 1:10 glass to waste volume ratio (37 up to 88 wt%) on the GCM wasteform microstructure and leaching properties were analysed. The open porosity ranged between 0.84 and {approx}13.2 % for the highest waste load. Significant changes occurred in the microstructure, phases present and wasteform durability at different waste loading. At waste loading up to 73 wt% of spent clinoptilolite, the GCM microstructure consists of several crystalline phases (clinoptilolite, sodalite, wollastonite and CsCl) that were fully encapsulated by a glass matrix. This leads to a low normalized leaching rate of Cs (remaining below 6.35 x 10{sup -6} g/cm{sup 2} day in a GCM with 73 wt% waste) during a leaching test for 7 days conducted using ASTM C1220-98. In GCM's with waste loading exceeding 73 wt%, the crystalline phases present (clinoptilolite and CsCl) were not fully encapsulated by the glass matrix hence the normalized leaching rate of Cs was as high as 9.06 x 10{sup -4} g/cm{sup 2} day at waste loading of {>=}80 wt%.

  12. LLW disposal wasteform preparation in the UK: the role of high force compaction

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, L. F.; Fearnley, I. G. [British Nuclear Fuels Ltd., Sellafield (United Kingdom)

    1991-07-01

    British Nuclear Fuels plc (BNFL) owns and operates the principal UK solid low level radioactive waste (LLW) disposal site. The site is located at Drigg in West Cumbria some 6 km to the south east of BNFL's Sellafield reprocessing complex. Sellafield is the major UK generator of LLW, accounting for about 85% of estimated future arisings of raw (untreated, unpackaged) waste. Non-Sellafield consignors to the Drigg site include other BNFL production establishments, nuclear power stations, sites of UKAEA, Ministry of Defence facilities, hospitals, universities, radioisotope production sites and various other industrial organisations. In September 1987, BNFL announced a major upgrade of operations at the Drigg site aimed at improving management practices, the efficiency of space utilisation and enhancing the visual impact of disposal operations. During 1989 a review of plans for compaction and containerisation of Sellafield waste identified that residual voidage in ISO freight containers could be significant even after the introduction of compaction. Subsequent studies which examined a range of compaction and packaging options concluded that the preferred scheme centred on the use of high force compaction (HFC) of compactable waste, and grouting to take up readily accessible voidage in the wasteform. The paper describes the emergence of high force compaction as the preferred scheme for wasteform preparation and subsequent benefits against the background of the overall development of Low Level Waste disposal operations at Drigg.

  13. Effects of gamma-ray irradiation on leaching of simulated {sup 133}Cs{sup +} radionuclides from geopolymer wasteforms

    Energy Technology Data Exchange (ETDEWEB)

    Deng, Ning; An, Hao; Cui, Hao, E-mail: cuihao@nju.edu.cn; Pan, Yang; Wang, Bing; Mao, Linqiang; Zhai, Jianping

    2015-04-15

    Highlights: • γ-ray irradiation caused more Cs{sup +} leaching out from geopolymer wasteform. • Pore structure change induced by irradiation caused the increase of leachability. • Fly-ash-based geopolymer is a potential material for radionuclide immobilization. - Abstract: Leaching of simulated {sup 133}Cs{sup +} radionuclides from geopolymer wasteforms was examined with regard to effects from gamma-ray irradiation. Specifically, the compressive strengths, microstructures, pore structures, and leaching resistance of geopolymer wasteforms before and after irradiation were characterized. The leaching experiments were performed by immersion of wasteforms in deionized water, ground water, and seawater. It was found that gamma rays did not produce significant morphological changes, except for changes in the pore size distribution. The cumulative leaching fraction of all the leachants from the irradiated samples increased relative to the non-radiated samples, particularly during long leaching periods (11–42 days). These results, and those from a mercury intrusion porosimeter analysis, can be attributed to irradiation-induced changes in pore structure. All the leaching indexes were greater than the minimum acceptable value of 6.0 set by the American Nuclear Society Standards committee, which indicated that the fly-ash geopolymers are suitable for radionuclide immobilization. However, the effects of gamma-ray irradiation on the immobilization of radionuclides cannot be ignored.

  14. Instrumentation for studying binder burnout in an immobilized plutonium ceramic wasteform

    Energy Technology Data Exchange (ETDEWEB)

    Mitchell, M; Pugh, D; Herman, C

    2000-04-21

    The Plutonium Immobilization Program produces a ceramic wasteform that utilizes organic binders. Several techniques and instruments were developed to study binder burnout on full size ceramic samples in a production environment. This approach provides a method for developing process parameters on production scale to optimize throughput, product quality, offgas behavior, and plant emissions. These instruments allow for offgas analysis, large-scale TGA, product quality observation, and thermal modeling. Using these tools, results from lab-scale techniques such as laser dilametry studies and traditional TGA/DTA analysis can be integrated. Often, the sintering step of a ceramification process is the limiting process step that controls the production throughput. Therefore, optimization of sintering behavior is important for overall process success. Furthermore, the capabilities of this instrumentation allows better understanding of plant emissions of key gases: volatile organic compounds (VOCs), volatile inorganics including some halide compounds, NO{sub x}, SO{sub x}, carbon dioxide, and carbon monoxide.

  15. Release of uranium and thorium from monazite ores: implications for phosphatic wasteforms

    Energy Technology Data Exchange (ETDEWEB)

    Read, D. [Reading Univ. (United Kingdom); Jarvis, N.V.; Andreoli, M.A.G. [Atomic Energy Corp. of South Africa Ltd., Pretoria (South Africa)

    1998-12-31

    Monazite is an important economic source of uranium, thorium, and the rare earth elements. Economically viable concentrations occur in placer deposits and as massive ore bodies in igneous and metamorphic provinces throughout the world. This and its refractory nature has led to speculation that artificial phosphate-based matrices similar in composition to monazite may prove useful as wasteforms for high level radioactive waste. Before a detailed evaluation can be made, however, it is worthwhile to assess the long term degradation behaviour of natural monazites in diverse hydrogeochemical environments. This paper focusses on the Steenkampskraal monazite mine, South Africa and compares findings with those from monazite occurrences in the Palaeozoic rocks of Britain and Western France. In all cases where the monazite has altered, a marked fractionation of the lanthanide series elements is apparent, together with substantial redistribution of uranium and thorium. The implications for waste encapsulation are discussed in terms of the potential for groundwater transport away from the source. (orig.)

  16. The influence of glass composition on crystalline phase stability in glass-ceramic wasteforms

    Energy Technology Data Exchange (ETDEWEB)

    Maddrell, Ewan, E-mail: ewan.r.maddrell@nnl.co.uk [National Nuclear Laboratory, Sellafield, Seascale, Cumbria CA20 1PG (United Kingdom); Thornber, Stephanie; Hyatt, Neil C. [Department of Materials Science and Engineering, The University of Sheffield, Mappin Street, Sheffield S1 3JD (United Kingdom)

    2015-01-15

    Highlights: • Crystalline phase formation shown to depend on glass matrix composition. • Zirconolite forms as the sole crystalline phase only for most aluminous glasses. • Thermodynamics indicate that low silica activity glasses stabilise zirconolite. - Abstract: Zirconolite glass-ceramic wasteforms were prepared using a suite of Na{sub 2}O–Al{sub 2}O{sub 3}–B{sub 2}O{sub 3}–SiO{sub 2} glass matrices with variable Al:B ratios. Zirconolite was the dominant crystalline phase only for the most alumina rich glass compositions. As the Al:B ratio decreased zirconolite was replaced by sphene, zircon and rutile. Thermodynamic data were used to calculate a silica activity in the glass melt below which zirconolite is the favoured crystalline phase. The concept of the crystalline reference state of glass melts is then utilised to provide a physical basis for why silica activity varies with the Al:B ratio.

  17. The effects of {gamma}-radiation on model vitreous wasteforms intended for the disposal of intermediate and high level radioactive wastes in the United Kingdom

    Energy Technology Data Exchange (ETDEWEB)

    McGann, O.J.; Bingham, P.A.; Hand, R.J.; Gandy, A.S. [Immobilisation Science Laboratory, Department of Materials Science and Engineering, University of Sheffield, Sir Robert Hadfield Building, Mappin Street, Sheffield S1 3JD (United Kingdom); Kavcic, M.; Zitnik, M.; Bucar, K. [J. Stefan Institute, Jamova 39, SI-1000 Ljubljana (Slovenia); Edge, R. [Dalton Cumbrian Facility, University of Manchester, Westlakes Science and Technology Park, Moor Row, Cumbria CA24 3HA (United Kingdom); Hyatt, N.C., E-mail: n.c.hyatt@sheffield.ac.uk [Immobilisation Science Laboratory, Department of Materials Science and Engineering, University of Sheffield, Sir Robert Hadfield Building, Mappin Street, Sheffield S1 3JD (United Kingdom)

    2012-10-15

    The effect of {gamma}-radiation on a variety of model vitreous wasteforms applied to, or conceived for, immobilisation of UK intermediate and high level radioactive wastes was studied up to a dose of 8 MGy. It was determined that {gamma}-irradiation up to this dose had no significant effect upon the mechanical properties of the wasteforms and there was no evidence of residual structural defects. FTIR and Raman spectroscopy showed no evidence of radiation directly affecting the silicate network of the glasses. The negligible impact of this {gamma}-irradiation dose on the physical properties of the glass was attributed to the presence of multivalent ions, particularly Fe, and a mechanism by which the electron-hole pairs generated by {gamma}-irradiation were annihilated by the Fe{sup 2+}-Fe{sup 3+} redox mechanism. However, reduction of sulphur species in response to {gamma}-radiation was demonstrated by S K-edge XANES and XES data.

  18. Calcium-borosilicate glass-ceramics wasteforms to immobilize rare-earth oxide wastes from pyro-processing

    Science.gov (United States)

    Kim, Miae; Heo, Jong

    2015-12-01

    Glass-ceramics containing calcium neodymium(cerium) oxide silicate [Ca2Nd8-xCex(SiO4)6O2] crystals were fabricated for the immobilization of radioactive wastes that contain large portions of rare-earth ions. Controlled crystallization of alkali borosilicate glasses by heating at T ≥ 750 °C for 3 h formed hexagonal Ca-silicate crystals. Maximum lanthanide oxide waste loading was >26.8 wt.%. Ce and Nd ions were highly partitioned inside Ca-silicate crystals compared to the glass matrix; the rare-earth wastes are efficiently immobilized inside the crystalline phases. The concentrations of Ce and Nd ions released in a material characterization center-type 1 test were below the detection limit (0.1 ppb) of inductively coupled plasma mass spectroscopy. Normalized release values performed by a product consistency test were 2.64·10-6 g m-2 for Ce ion and 2.19·10-6 g m-2 for Nd ion. Results suggest that glass-ceramics containing calcium neodymium(cerium) silicate crystals are good candidate wasteforms for immobilization of lanthanide wastes generated by pyro-processing.

  19. State of knowledge on the water radiolysis in cemented wasteforms and its approach by simulation; Etat des connaissances sur la radiolyse de l'eau dans les colis de dechets cimentes et son approche par simulation

    Energy Technology Data Exchange (ETDEWEB)

    Bouniol, P

    2004-07-01

    The decomposition of water under radiation within the cementitious matrix is at the origin of a potential source of harmful effects in the wasteform and their environment (pressurization and emanation of di-hydrogen) which can have an impact on the safety. In the aim of a better evaluation of the 'H{sub 2}' risk induced by such a complex and heterogeneous system, this document is an analysis of the elements necessary for a global understanding of the radiolysis in the cemented wasteform to be achieved: - summary of the basic knowledge on water radiolysis with transposition to the cementitious medium, - critical review of the various phenomenologies at work in a wasteform (radioactive source-term, gas transport, mineral equilibria); description of their mutual couplings and of their feedback on radiolytic chemistry; identification of the determining parameters, - presentation of a selection of experimental facts putting in light some theoretical points, - presentation of an outline of operational model deriving from the global vision; presentation of an adapted tool for simulation (CHEMSIMUL) and study of the influence of the principal parameters, starting from a reference case. The main result of this work is that it is shown, in the case of a {beta}{gamma} source term, that the control of the pore fluid composition by calcium octo-hydrate peroxide constitutes an efficient regulating mechanism for the radiolysis and H{sub 2} production. Not likely possible in the case of an {alpha} source term, this suggests a separate management of the wasteform according to their radiological contents. The gaps and limits of the model which are also evoked are promising of a lot of research prospects, primarily of a fundamental nature (impact of the porous medium). (author)

  20. Al2O3对独居石玻璃陶瓷固化体的影响%Effect of Al2O3 on Monazite Glass-ceramic Wasteform

    Institute of Scientific and Technical Information of China (English)

    廖其龙; 廖春娟; 向光华; 潘社奇; 牟涛

    2014-01-01

    本文研究了Al2O3掺量对独居石玻璃陶瓷固化体结构和化学稳定性的影响。用傅里叶变换红外光谱(FTIR)和X射线衍射(XRD)方法表征样品结构,用溶解速率法和全谱直读等离子体发射光谱(ICP-OES)分别测定样品在浸出液中浸泡后的失重速率及各元素的浸出浓度,以研究固化体的化学稳定性。研究结果表明:当Al2O3掺量为4%(摩尔分数)时,在980℃下保温3h得到的独居石玻璃陶瓷固化体具有较高的化学稳定性,浸泡14d时其质量浸出率最低,约为8.1ng/(cm2·min),其中Ce、La元素在浸出液中均未检出;固化体的主晶相为独居石,结构中含有大量稳定的正磷酸基团[PO4]3-和少量的焦磷酸基团[P2O7]4-,不存在偏磷酸基团[PO3]-。%The effects of monazite glass-ceramic wasteforms containing different Al2O3 contents on their structures and properties were investigated .The structure of the glass-ceramic wasteforms was analyzed by Fourier transform infrared (FTIR) and X-radiation diffraction (XRD ) . The chemical stability of monazite glass-ceramic wasteforms was measured by dissolution rate and inductively coupled plasma optical emission spectrome-try (ICP-OES) method .The results show that the chemical stability of monazite glass-ceramic wasteforms with 4% (mole fraction) Al2O3 and made at 980 ℃ for 3 h is opti-mal .T he 14 d leaching rate of monazite glass-ceramic wasteforms is about 8.1 ng/(cm2 · min) ,which is the lowest in all the samples .The main crystalline phase of the as-prepared glass-ceramic wasteforms is monazite .There are a large number of [PO4 ]3 -groups ,a small number of [P2O7 ]4 - groups ,and no [PO3 ]- groups in the wasteforms .

  1. The effect of pre-treatment parameters on the quality of glass-ceramic wasteforms for plutonium immobilisation, consolidated by hot isostatic pressing

    Science.gov (United States)

    Thornber, Stephanie M.; Heath, Paul G.; Da Costa, Gabriel P.; Stennett, Martin C.; Hyatt, Neil C.

    2017-03-01

    Glass-ceramics with high glass fractions (70 wt%) were fabricated in stainless steel canisters by hot isostatic pressing (HIP), at laboratory scale. High (600 °C) and low (300 °C) temperature pre-treatments were investigated to reduce the canister evacuation time and to understand the effect on the phase assemblage and microstructure of the hot isostatically pressed product. Characterisation of the HIPed materials was performed using scanning electron microscopy (SEM), coupled with energy dispersive X-ray analysis (EDX) and powder X-ray diffraction (XRD). This analysis showed the microstructure and phase assemblage was independent of the variation in pre-treatment parameters. It was demonstrated that a high temperature pre-treatment of batch reagents, prior to the HIP cycle, is beneficial when using oxide precursors, in order to remove volatiles and achieve high quality dense materials. Sample throughput can be increased significantly by utilising a high temperature ex-situ calcination prior to the HIP cycle. Investigation of glass-ceramic wasteform processing utilising a glass frit precursor, produced a phase assemblage and microstructure comparable to that obtained using oxide precursors. The use of a glass frit precursor should allow optimised throughput of waste packages in a production facility, avoiding the need for a calcination pre-treatment required to remove volatiles from oxide precursors.

  2. Miscellaneous Waste-Form FEPs

    Energy Technology Data Exchange (ETDEWEB)

    A. Schenker

    2000-12-08

    The US DOE must provide a reasonable assurance that the performance objectives for the Yucca Mountain Project (YMP) potential radioactive-waste repository can be achieved for a 10,000-year post-closure period. The guidance that mandates this direction is under the provisions of 10 CFR Part 63 and the US Department of Energy's ''Revised Interim Guidance Pending Issuance of New US Nuclear Regulatory Commission (NRC) Regulations (Revision 01, July 22, 1999), for Yucca Mountain, Nevada'' (Dyer 1999 and herein referred to as DOE's Interim Guidance). This assurance must be demonstrated in the form of a performance assessment that: (1) identifies the features, events, and processes (FEPs) that might affect the performance of the potential geologic repository; (2) examines the effects of such FEPs on the performance of the potential geologic repository; (3) estimates the expected annual dose to a specified receptor group; and (4) provides the technical basis for inclusion or exclusion of specific FEPs.

  3. Simulated monazite crystalline wasteform La0.4Nd0.1Y0.1Gd0.1Sm0.1Ce0.1Ca0.1(P0.9Mo0.1O4): Synthesis, phase stability and chemical durability study

    Science.gov (United States)

    Pratheep Kumar, Sathasivam; Gopal, Buvaneswari

    2015-03-01

    In this work, incorporation of hexavalent molybdenum and selected trivalent lanthanides using divalent calcium as charge compensator into the monazite structure were studied. Rare earth substituted phosphomolybdates of the formula REE0.9Ca0.1P0.9Mo0.1O4 (REE = Ce, Nd, Sm, Gd) and the wasteform La0.4Nd0.1Y0.1Gd0.1Sm0.1Ce0.1Ca0.1P0.9Mo0.1O4 were synthesized by simple solution route. The prepared compounds were characterized by powder X-ray diffraction, Fourier transformed infrared spectra, thermogravimetric analysis, energy dispersive X-ray analysis and scanning electron microscopic techniques. Chemical durability of the wasteform was studied by dynamic MCC-5 test for a period of one month. Normalized elemental mass loss and leach rate of molybdenum was found to be in the order of 103 g/m2 and 103-101 g/m2/d respectively. Polymer-monazite composite wasteform was prepared to control the leaching of molybdenum. The composite approach reduced molybdenum leach rate order from 101 to 10-4 g/m2/d.

  4. Effect of composition variations on the long-term wasteform behavior of vitrified domestic waste incineration fly-ash purification residues; Influence des variations de composition des vitrifiats de refiom - residus d'epuration des fumees d'incineration d'ordures menageres - sur leur comportement a long terme

    Energy Technology Data Exchange (ETDEWEB)

    Frugier, P.

    2000-07-01

    The effect of variations in the composition of fly-ash purification residue from incinerated domestic waste on the quality of the containment achieved by vitrification was investigated. Three main factors determine the long-term containment quality: the production of a vitrified wasteform, the occurrence of possible crystallization, and the key parameters of long-term alteration in aqueous media. Each of these aspects is described within a composition range defined by variations in the three major elements. (silicon, calcium and aluminum) and two groups of constituents (alkali metals and toxic elements). The silicon fraction in the fly-ash residue was found to be decisive: it is impossible to obtain a satisfactory vitrified wasteform below a given silicon concentration. Compounds with the lowest silica content also exhibited the greatest tendency to crystallize under the cooling conditions prevailing in industrial processes (the dominant crystallized phase is a melilite that occupies a significant fraction of the material and considerably modifies the alteration mechanisms). The initial alteration rate in pure water and the altered glass thickness measured in a closed system at an advanced stage of the dissolution reaction are both inversely related to the silicon concentration in the glass. Several types of long-term behavior were identified according to the composition range, the process conditions and the vitrified waste disposal scenario. Four distinct 'classes' of vitrified wasteform were defined for direct application in industrial processes. (author)

  5. Simulated monazite crystalline wasteform La{sub 0.4}Nd{sub 0.1}Y{sub 0.1}Gd{sub 0.1}Sm{sub 0.1}Ce{sub 0.1}Ca{sub 0.1}(P{sub 0.9}Mo{sub 0.1}O{sub 4}): Synthesis, phase stability and chemical durability study

    Energy Technology Data Exchange (ETDEWEB)

    Pratheep Kumar, Sathasivam, E-mail: pratheepvit@gmail.com; Gopal, Buvaneswari

    2015-03-15

    Highlights: • Hexavalent molybdenum and selected lanthanides were immobilized into monazite structure at low temperature. • The wasteform has been characterized by powder XRD, FTIR, SEM-EDX and XPS analysis. • Immobilization of molybdenum was proved by energy dispersive X-ray analysis. • Chemical durability of the wasteform was studied by dynamic Soxhlet test. • Polymer–ceramic composite approach reduced the leaching of molybdenum from monazite lattice. - Abstract: In this work, incorporation of hexavalent molybdenum and selected trivalent lanthanides using divalent calcium as charge compensator into the monazite structure were studied. Rare earth substituted phosphomolybdates of the formula REE{sub 0.9}Ca{sub 0.1}P{sub 0.9}Mo{sub 0.1}O{sub 4} (REE = Ce, Nd, Sm, Gd) and the wasteform La{sub 0.4}Nd{sub 0.1}Y{sub 0.1}Gd{sub 0.1}Sm{sub 0.1}Ce{sub 0.1}Ca{sub 0.1}P{sub 0.9}Mo{sub 0.1}O{sub 4} were synthesized by simple solution route. The prepared compounds were characterized by powder X-ray diffraction, Fourier transformed infrared spectra, thermogravimetric analysis, energy dispersive X-ray analysis and scanning electron microscopic techniques. Chemical durability of the wasteform was studied by dynamic MCC-5 test for a period of one month. Normalized elemental mass loss and leach rate of molybdenum was found to be in the order of 10{sup 3} g/m{sup 2} and 10{sup 3}–10{sup 1} g/m{sup 2}/d respectively. Polymer-monazite composite wasteform was prepared to control the leaching of molybdenum. The composite approach reduced molybdenum leach rate order from 10{sup 1} to 10{sup −4} g/m{sup 2}/d.

  6. Release of uranium from candidate wasteforms

    OpenAIRE

    Collier, N.; Harrison, M.; Brogden, M,; Hanson, B

    2012-01-01

    Large volumes of depleted natural and low-enriched uranium exist in the UK waste inventory. This work reports on initial investigations of the leaching performance of candidate glass and cement encapsulation matrices containing UO3 powder as well as that of uranium oxide powders. The surface areas of UO3 powder and the monolith samples of UO3 conditioned in the glass and cement matrices were very different making leaching comparisons difficult. The results showed that for both types of monoli...

  7. Effect of Concrete Wasteform Properties on Radionuclide Migration

    Energy Technology Data Exchange (ETDEWEB)

    Wellman, Dawn M.; Bovaird, Chase C.; Mattigod, Shas V.; Parker, Kent E.; Ermi, Ruby M.; Wood, Marcus I.

    2008-09-30

    The objective of this investigation was to initiate numerous sets of concrete-soil half-cell tests to quantify 1) diffusion of I and Tc from concrete into uncontaminated soil after 1 and 2 years, 2) I and Re (set 1) and Tc (set 2) diffusion from fractured concrete into uncontaminated soil, and 3) evaluate the moisture distribution profile within the sediment half-cell. These half-cells will be section in FY2009 and FY2010. Additionally, 1) concrete-soil half-cells initiated during FY2007 using fractured prepared with and without metallic iron, half of which were carbonated using carbonated, were sectioned to evaluate the diffusion of I and Re in the concrete part of the half-cell under unsaturated conditions (4%, 7%, and 15% by wt moisture content), 2) concrete-soil half cells containing Tc were sectioned to measure the diffusion profile in the soil half-cell unsaturated conditions (4%, 7%, and 15% by wt moisture content), and 3) solubility measurements of uranium solid phases were completed under concrete porewater conditions. The results of these tests are presented.

  8. Timing of Getter Material Addition in Cementitious Wasteforms

    Science.gov (United States)

    Lawter, A.; Qafoku, N. P.; Asmussen, M.; Neeway, J.; Smith, G. L.

    2015-12-01

    A cementitious waste form, Cast Stone, is being evaluated as a possible supplemental immobilization technology for the Hanford sites's low activity waste (LAW), which contains radioactive 99Tc and 129I, as part of the tank waste cleanup mission. Cast Stone is made of a dry blend 47% blast furnace slag, 45% fly ash, and 8% ordinary Portland cement, mixed with a low-activity waste (LAW). To improve the retention of Tc and/or I in Cast Stone, materials with a high affinity for Tc and/or I, termed "getters," can be added to provide a stable domain for the radionuclides of concern. Previous testing conducted with a variety of getters has identified Tin(II)-Apatite and Silver Exchanged Zeolite as promising candidates for Tc and I, respectively. Investigation into the sequence in which getters are added to Cast Stone was performed following two methods: 1) adding getters to the Cast Stone dry blend, and then mixing with liquid waste, and 2) adding getters to the liquid waste first, followed by addition of the Cast Stone dry blend. Cast Stone monolith samples were prepared with each method and leach tests, following EPA method 1315, were conducted in either distilled water or simulated vadose zone porewater for a period of up to 63 days. The leachate was analyzed for Tc, I, Na, NO3-, NO2- and Cr with ICP-MS, ICP-OES and ion chromatography and the results indicated that the Cast Stone with getter addition in the dry blend mix (method 1) has lower rates of Tc and I leaching. The mechanisms of radionuclide release from the Cast Stone were also investigated with a variety of solid phase characterization techniques of the monoliths before and after leaching, such as XRD, SEM/EDS, TEM/SAED and other spectroscopic techniques.

  9. Waste-Form Development Program. Annual progress report, October 1981-September 1982

    Energy Technology Data Exchange (ETDEWEB)

    Neilson, R.M. Jr.; Colombo, P.

    1982-09-01

    Low-level wastes (LLW) at nuclear facilities have traditionally been solidified using portland cement (with and without additives). Urea-formaldehyde has been used for LLW solidification while bitumen (asphalt) and thermosetting polymers will be applied to domestic wastes in the near future. Operational difficulties have been observed with each of these solidification agents. Such difficulties include incompatibility with waste constitutents inhibiting solidification, premature setting, free standing water and fires. Some specific waste types have proven difficult to solidify with one or more of the contemporary agents. Similar problems are also anticipated for the solidification of new wastes, which are generated using advanced volume reduction technologies, and with the application of additional agents which may be introduced in the near future for the solidification of LLW. In the Waste Form Development program, contemporary solidification agents are being investigated relative to their potential applications to major fuel cycle and non-fuel cycle LLW streams. The range of conditions under which these solidification agents can be satisfactorily applied to specific LLW streams is being determined. These studies are primarily directed towards defining operating parameters for both improved solidification of problem wastes such as ion exchange resins, organic liquids and oils for which prevailing processes, as currently employed, appear to be inadequate, and solidification of new LLW streams including high solids content evaporator concentrates, dry solids, and incinerator ash generated from advanced volume reduction technologies. Solidified waste forms are tested and evaluated to demonstrate compliance with waste form performance and shallow land burial (SLB) acceptance criteria and transportation requirements (both as they currently exist and as they are anticipated to be modified with time).

  10. Solid Solubilities of Pu, U, Gd and Hf in Candidate Ceramic Nuclear Wasteforms

    Energy Technology Data Exchange (ETDEWEB)

    Vance, Eric R.; Carter, M. L.; Lumpkin, G. R.; Day, R. A.; Begg, B. D.

    2001-04-02

    This goal of this research project was to determine the solid solubility of Pu, U, Gd, and Hf in candidate ceramics for immobilization of high-level nuclear waste. The experimental approach was to saturate each phase by adding more than the solid solubility limit of the given cation, using a nominated substitution scheme, and then analyzing the candidate phase that formed to evaluate the solid solubility limit under firing conditions. Confirmation that the solid solution limit had been reached insofar as other phases rich in the cation of interest was also required. The candidate phases were monazite, titanite, zirconolite, perovskite, apatite, pyrochlore, and brannerite. The valences of Pu and U were typically deduced from the firing atmosphere, and charge balancing in the candidate phase composition as evaluated from electron microscopy, although in some cases it was measured directly by x-ray absorption and diffuse reflectance spectroscopies (for U). Tetravalent Pu and U have restricted (< 0.1 formula units) solid solubility in apatite, titanite, and perovskite. Trivalent Pu has a larger solubility in apatite and perovskite than Pu4+. U3+ appears to be a credible species in reduced perovskite with a solubility of {approximately} 0.25 f.u. as opposed to {approximately} 0.05 f.u. for U4+. Pu4+ is a viable species in monazite and is promoted at lower firing temperatures ({approximately} 800 C) in an air atmosphere. Hf solubility is restricted in apatite, monazite (< 0.1 f.u.), but is {approximately} 0.2 and 0.5 f.u. in brannerite and titanite, respectively. Gd solubility is extended in all phases except for titanite ({approximately} 0.3 f.u.). U5+ was identified by DRS observations of absorption bands in the visible/near infrared photon energy ranges in brannerite and zirconolite, and U4+ in zirconolite was similarly identified.

  11. Radionuclide Retention Mechanisms in Secondary Waste-Form Testing: Phase II

    Energy Technology Data Exchange (ETDEWEB)

    Um, Wooyong; Valenta, Michelle M.; Chung, Chul-Woo; Yang, Jungseok; Engelhard, Mark H.; Serne, R. Jeffrey; Parker, Kent E.; Wang, Guohui; Cantrell, Kirk J.; Westsik, Joseph H.

    2011-09-26

    This report describes the results from laboratory tests performed at Pacific Northwest National Laboratory (PNNL) for Washington River Protection Solutions (WRPS) to evaluate candidate stabilization technologies that have the potential to successfully treat liquid secondary waste stream effluents produced by the Hanford Tank Waste Treatment and Immobilization Plant (WTP). WRPS is considering the design and construction of a Solidification Treatment Unit (STU) for the Effluent Treatment Facility (ETF) at Hanford. The ETF, a multi-waste, treatment-and-storage unit that has been permitted under the Resource Conservation and Recovery Act (RCRA), can accept dangerous, low-level, and mixed wastewaters for treatment. The STU needs to be operational by 2018 to receive secondary liquid waste generated during operation of the WTP. The STU will provide the additional capacity needed for ETF to process the increased volume of secondary waste expected to be produced by WTP. This report on radionuclide retention mechanisms describes the testing and characterization results that improve understanding of radionuclide retention mechanisms, especially for pertechnetate, {sup 99}TcO{sub 4}{sup -} in four different waste forms: Cast Stone, DuraLith alkali aluminosilicate geopolymer, encapsulated fluidized bed steam reforming (FBSR) product, and Ceramicrete phosphate bonded ceramic. These data and results will be used to fill existing data gaps on the candidate technologies to support a decision-making process that will identify a subset of the candidate waste forms that are most promising and should undergo further performance testing.

  12. Waste-Form Development Program. Annual progress report, October 1981-September 1982

    Energy Technology Data Exchange (ETDEWEB)

    Neilson, R.M. Jr.; Colombo, P.

    1982-09-01

    Low-level wastes (LLW) at nuclear facilities have traditionally been solidified using portland cement (with and without additives). Urea-formaldehyde has been used for LLW solidification while bitumen (asphalt) and thermosetting polymers will be applied to domestic wastes in the near future. Operational difficulties have been observed with each of these solidification agents. Such difficulties include incompatibility with waste constitutents inhibiting solidification, premature setting, free standing water and fires. Some specific waste types have proven difficult to solidify with one or more of the contemporary agents. Similar problems are also anticipated for the solidification of new wastes, which are generated using advanced volume reduction technologies, and with the application of additional agents which may be introduced in the near future for the solidification of LLW. In the Waste Form Development program, contemporary solidification agents are being investigated relative to their potential applications to major fuel cycle and non-fuel cycle LLW streams. The range of conditions under which these solidification agents can be satisfactorily applied to specific LLW streams is being determined. These studies are primarily directed towards defining operating parameters for both improved solidification of problem wastes such as ion exchange resins, organic liquids and oils for which prevailing processes, as currently employed, appear to be inadequate, and solidification of new LLW streams including high solids content evaporator concentrates, dry solids, and incinerator ash generated from advanced volume reduction technologies. Solidified waste forms are tested and evaluated to demonstrate compliance with waste form performance and shallow land burial (SLB) acceptance criteria and transportation requirements (both as they currently exist and as they are anticipated to be modified with time).

  13. Solid Solubilities of Pu, U, Gd and Hf in Candidate Ceramic Nuclear Wasteforms

    Energy Technology Data Exchange (ETDEWEB)

    Vance, Eric R.; Carter, M. L.; Lumpkin, G. R.; Day, R. A.; Begg, B. D.

    2001-04-02

    This goal of this research project was to determine the solid solubility of Pu, U, Gd, and Hf in candidate ceramics for immobilization of high-level nuclear waste. The experimental approach was to saturate each phase by adding more than the solid solubility limit of the given cation, using a nominated substitution scheme, and then analyzing the candidate phase that formed to evaluate the solid solubility limit under firing conditions. Confirmation that the solid solution limit had been reached insofar as other phases rich in the cation of interest was also required. The candidate phases were monazite, titanite, zirconolite, perovskite, apatite, pyrochlore, and brannerite. The valences of Pu and U were typically deduced from the firing atmosphere, and charge balancing in the candidate phase composition as evaluated from electron microscopy, although in some cases it was measured directly by x-ray absorption and diffuse reflectance spectroscopies (for U). Tetravalent Pu and U have restricted (< 0.1 formula units) solid solubility in apatite, titanite, and perovskite. Trivalent Pu has a larger solubility in apatite and perovskite than Pu4+. U3+ appears to be a credible species in reduced perovskite with a solubility of {approximately} 0.25 f.u. as opposed to {approximately} 0.05 f.u. for U4+. Pu4+ is a viable species in monazite and is promoted at lower firing temperatures ({approximately} 800 C) in an air atmosphere. Hf solubility is restricted in apatite, monazite (< 0.1 f.u.), but is {approximately} 0.2 and 0.5 f.u. in brannerite and titanite, respectively. Gd solubility is extended in all phases except for titanite ({approximately} 0.3 f.u.). U5+ was identified by DRS observations of absorption bands in the visible/near infrared photon energy ranges in brannerite and zirconolite, and U4+ in zirconolite was similarly identified.

  14. Hanford waste-form release and sediment interaction: A status report with rationale and recommendations for additional studies

    Energy Technology Data Exchange (ETDEWEB)

    Serne, R.J. (Pacific Northwest Lab., Richland, WA (USA)); Wood, M.I. (Westinghouse Hanford Co., Richland, WA (USA))

    1990-05-01

    This report documents the currently available geochemical data base for release and retardation for actual Hanford Site materials (wastes and/or sediments). The report also recommends specific laboratory tests and presents the rationale for the recommendations. The purpose of this document is threefold: to summarize currently available information, to provide a strategy for generating additional data, and to provide recommendations on specific data collection methods and tests matrices. This report outlines a data collection approach that relies on feedback from performance analyses to ascertain when adequate data have been collected. The data collection scheme emphasizes laboratory testing based on empiricism. 196 refs., 4 figs., 36 tabs.

  15. CBP [TASK 12] experimental study of the concrete salstone two-layer system

    Energy Technology Data Exchange (ETDEWEB)

    Samson, Eric [SIMCO Technologies, Inc., Ville de Québec, QC (Canada); Protiere, Yannick [SIMCO Technologies, Inc., Ville de Québec, QC (Canada)

    2016-11-01

    This report presents the results of a study which intended to study the behavior of concrete samples placed in contact with a wasteform mixture bearing high level of sulfate in its pore solution. A setup was prepared which consisted in a wasteform poured on top of vault concrete mixes (identified as Vault 1/4 and Vault 2 mixes) cured for approximately 6 months.

  16. Effect of ZnO and CaO on Alkali Borosilicate Glass Waste-form Immobilizing Simulated Mixed HLW%ZnO 和 CaO对模拟高放废液硅酸盐玻璃固化体性能的影响研究

    Institute of Scientific and Technical Information of China (English)

    张华; N.C.Hyatt; J.R.Stevens; R.Hand

    2015-01-01

    针对有些高放废液含有较多Fe、Cr、Ni过渡金属元素,在玻璃固化工艺过程中易于形成晶体,导致熔融玻璃体的黏度增加、化学稳定性变差以及工艺过程中易出现出料口堵塞等问题,研究了废物包容量为15%和20%、添加ZnO (5.6%)和CaO (1.75%)的配方对形成的4种玻璃固化体的物理性能(密度、硬度、断裂韧性)、化学性能(产品一致性测试和蒸汽腐蚀测试)和结构(X射线衍射析晶分析、拉曼光谱分析)的影响。研究分析显示,提高废物包容量至20%以及添加ZnO和CaO均可促进硼硅酸盐玻璃固化体网络结构的稳定性和化学稳定性,并增强玻璃体的密度,提高硬度;但玻璃固化体的高温黏度升高,断裂韧性下降。%Since the transit metals ,such as Fe ,Cr and Ni ,contained in some kinds of mixed HLW ,can likely to form crystal ,increase the melt viscosity ,destroy the chemi‐cal durability and block the discharge port .T he results obtained from investigating four glass waste‐forms ,including the alkali borosilicate glass matrix and alkali borosilicate glass matrix doped with 5.6% ZnO and 1.75% CaO in base matrixes ,immobilizing the simulated mixed HLW with 15% and 20% waste loadings aiming to determinate the effect of ZnO on the alkali borosilicate glass chemical durability with waste loading increasing ,were presented in this paper .Glass samples were characterized with XRD and Raman spectroscopy .The chemical durability was investigated using the standard protocols PCT and VHT .The XRD analysis results show that spinel crystal appears and grows in glass samples at the waste loading in 20% without ZnO addition and waste loading in 15% and 20% added ZnO .T he Raman spectroscopy analysis results indicate that ZnO and CaO can enhance the glass network connective ,and the chemical durability test results display that the addition of ZnO and CaO can improve the short term chemi‐cal durability of the glass samples ,except Zn20 obtained the lower value in VHT result , which is caused by the higher crystal ratio and lower fracture toughness .

  17. Synthesis of apatite and monazite waste form for immobilization of rare earth oxide radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, B. G.; Park, H. S.; Kim, I. T.; Lee, H. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-08-15

    In order to fabricate a monolithic waste form containing RE oxides, a vitrification at a high temperature or a ceramization by a HIP method is required. In this study, a series of monolithic wasteform with high waste loading were successfully produced at a mild condition, where the chemical structure was equivalent to the product by a high temperature process or a monolithic wasteform consisting of a durable ceramic host matrix for immobilizing RE elements.

  18. Radioactive Waste Conditioning, Immobilisation, And Encapsulation Processes And Technologies: Overview And Advances (Chapter 7)

    Energy Technology Data Exchange (ETDEWEB)

    Jantzen, Carol M. [Savannah River National Lab., Aiken SC (United States); Lee, William E. [Imperial College, London (United Kingdom). Dept. of Materials; Ojovan, Michael I. [Univ. of Sheffield (United Kingdom). Dept. of Materials Science and Engineering

    2012-10-19

    The main immobilization technologies that are available commercially and have been demonstrated to be viable are cementation, bituminization, and vitrification. Vitrification is currently the most widely used technology for the treatment of high level radioactive wastes (HLW) throughout the world. Most of the nations that have generated HLW are immobilizing in either alkali borosilicate glass or alkali aluminophosphate glass. The exact compositions of nuclear waste glasses are tailored for easy preparation and melting, avoidance of glass-in-glass phase separation, avoidance of uncontrolled crystallization, and acceptable chemical durability, e.g., leach resistance. Glass has also been used to stabilize a variety of low level wastes (LLW) and mixed (radioactive and hazardous) low level wastes (MLLW) from other sources such as fuel rod cladding/decladding processes, chemical separations, radioactive sources, radioactive mill tailings, contaminated soils, medical research applications, and other commercial processes. The sources of radioactive waste generation are captured in other chapters in this book regarding the individual practices in various countries (legacy wastes, currently generated wastes, and future waste generation). Future waste generation is primarily driven by interest in sources of clean energy and this has led to an increased interest in advanced nuclear power production. The development of advanced wasteforms is a necessary component of the new nuclear power plant (NPP) flowsheets. Therefore, advanced nuclear wasteforms are being designed for robust disposal strategies. A brief summary is given of existing and advanced wasteforms: glass, glass-ceramics, glass composite materials (GCM’s), and crystalline ceramic (mineral) wasteforms that chemically incorporate radionuclides and hazardous species atomically in their structure. Cementitious, geopolymer, bitumen, and other encapsulant wasteforms and composites that atomically bond and encapsulate

  19. Molecular Environmental Science Using Synchrotron Radiation: Chemistry and Physics of Waste Form Materials

    Energy Technology Data Exchange (ETDEWEB)

    Lindle, Dennis W.

    2011-04-21

    Production of defense-related nuclear materials has generated large volumes of complex chemical wastes containing a mixture of radionuclides. The disposition of these wastes requires conversion of the liquid and solid-phase components into durable, solid forms suitable for long-term immobilization. Specially formulated glass compositions and ceramics such as pyrochlores and apatites are the main candidates for these wastes. An important consideration linked to the durability of waste-form materials is the local structure around the waste components. Equally important is the local structure of constituents of the glass and ceramic host matrix. Knowledge of the structure in the waste-form host matrices is essential, prior to and subsequent to waste incorporation, to evaluate and develop improved waste-form compositions based on scientific considerations. This project used the soft-x-ray synchrotron-radiation-based technique of near-edge x-ray-absorption fine structure (NEXAFS) as a unique method for investigating oxidation states and structures of low-Z elemental constituents forming the backbones of glass and ceramic host matrices for waste-form materials. In addition, light metal ions in ceramic hosts, such as titanium, are also ideal for investigation by NEXAFS in the soft-x-ray region. Thus, one of the main objectives was to understand outstanding issues in waste-form science via NEXAFS investigations and to translate this understanding into better waste-form materials, followed by eventual capability to investigate “real” waste-form materials by the same methodology. We conducted several detailed structural investigations of both pyrochlore ceramic and borosilicate-glass materials during the project and developed improved capabilities at Beamline 6.3.1 of the Advanced Light Source (ALS) to perform the studies.

  20. Near field and altered zone environmental report Volume I: technical bases for EBS design

    Energy Technology Data Exchange (ETDEWEB)

    Wilder, D. G., LLNL

    1997-08-01

    This report presents an updated summary of results for the waste package (WP) and engineered barrier system (EBS) evaluations, including materials testing, waste-form characterization, EBS performance assessments, and near-field environment (NFE) characterization. Materials testing, design criteria and concept development, and waste-form characterization all require an understanding of the environmental conditions that will interact with the WP and EBS. The Near-Field Environment Report (NFER) was identified in the Waste Package Plan (WPP) (Harrison- Giesler, 1991) as the formal means for transmitting and documenting this information.

  1. Relevance of the choice of spark plasma sintering parameters in obtaining a suitable microstructure for iodine-bearing apatite designed for the conditioning of I-129

    Energy Technology Data Exchange (ETDEWEB)

    Campayo, L., E-mail: lionel.campayo@cea.fr [CEA, DEN, DTCD/SECM/LDMC – Marcoule, F-30207 Bagnols-sur-Cèze (France); Le Gallet, S. [Laboratoire Interdisciplinaire Carnot de Bourgogne, UMR 6303 CNRS-UB, 9 Av. Alain Savary, BP 47870, 21078 Dijon Cedex (France); Perret, D.; Courtois, E. [CEA, DEN, DTCD/SECM/LDMC – Marcoule, F-30207 Bagnols-sur-Cèze (France); Cau Dit Coumes, C. [CEA, DEN, DTCD/SPDE/LP2C – Marcoule, F-30207 Bagnols-sur-Cèze (France); Grin, Yu. [Max-Planck-Institut für Chemische Physik fester Stoffe, Nöthnitzer Straße 40, 01187 Dresden (Germany); Bernard, F. [Laboratoire Interdisciplinaire Carnot de Bourgogne, UMR 6303 CNRS-UB, 9 Av. Alain Savary, BP 47870, 21078 Dijon Cedex (France)

    2015-02-15

    Highlights: • Modeling of reactive sintering by SPS of an iodoapatite for waste immobilization. • Use of a statistical approach to surmount the complexity of the process. • The supposed most resistant microstructure towards leaching is obtained at 450 °C. • Pressure has no influence in the liquid sintering regime. - Abstract: The high chemical durability of iodine-bearing apatite phases makes them potentially attractive for immobilizing radioactive iodine. Reactive spark plasma sintering provides a dense ceramic as a wasteform. A design-of-experiments (DOE) approach was adopted to identify the main process/material parameters and their first order interactions in order to specify experimental conditions guaranteeing complete reaction, relative density of the wasteform exceeding 92% and the largest possible grain size. For a disposal of the wasteform in a deep geological repository, these characteristics allow minimization of the iodine release by contact with groundwater. It was found that sintering at a temperature of 450 °C with an initial specific surface area of 3.3 m{sup 2} g{sup −1} for the powder reactants is sufficient in itself to achieve the targeted characteristics of the wasteform. However, this relies on a liquid sintering regime the efficiency of which can be limited by the lead iodide initial content in the mix as well as by its particle size.

  2. Modeling the long-term durability of concrete barriers in the context of low-activity waste storage

    Directory of Open Access Journals (Sweden)

    Samson E.

    2013-07-01

    Full Text Available The paper investigates the long-term durability of concrete barriers in contact with a cementitious wasteform designed to immobilize low-activity nuclear waste. The high-pH pore solution of the wasteform contains high concentration level of sulfate, nitrate, nitrite and alkalis. The multilayer concrete/wasteform system was modeled using a multiionic reactive transport model accounting for coupling between species, dissolution/ precipitation reactions, and feedback effect. One of the primary objectives was to investigate the risk associated with the presence of sulfate in the wasteform on the durability of concrete. Simulation results showed that formation of expansive phases, such as gypsum and ettringite, into the concrete barrier was not extensive. Based on those results, it was not possible to conclude that concrete would be severely damaged, even after 5,000 years. Lab work was performed to provide data to validate the modeling results. Paste samples were immersed in sulfate contact solutions and analyzed to measure the impact of the aggressive environment on the material. The results obtained so far tend to confirm the numerical simulations.

  3. DUSTMS-D: DISPOSAL UNIT SOURCE TERM - MULTIPLE SPECIES - DISTRIBUTED FAILURE DATA INPUT GUIDE.

    Energy Technology Data Exchange (ETDEWEB)

    SULLIVAN, T.M.

    2006-01-01

    Performance assessment of a low-level waste (LLW) disposal facility begins with an estimation of the rate at which radionuclides migrate out of the facility (i.e., the source term). The focus of this work is to develop a methodology for calculating the source term. In general, the source term is influenced by the radionuclide inventory, the wasteforms and containers used to dispose of the inventory, and the physical processes that lead to release from the facility (fluid flow, container degradation, wasteform leaching, and radionuclide transport). Many of these physical processes are influenced by the design of the disposal facility (e.g., how the engineered barriers control infiltration of water). The complexity of the problem and the absence of appropriate data prevent development of an entirely mechanistic representation of radionuclide release from a disposal facility. Typically, a number of assumptions, based on knowledge of the disposal system, are used to simplify the problem. This has been done and the resulting models have been incorporated into the computer code DUST-MS (Disposal Unit Source Term-Multiple Species). The DUST-MS computer code is designed to model water flow, container degradation, release of contaminants from the wasteform to the contacting solution and transport through the subsurface media. Water flow through the facility over time is modeled using tabular input. Container degradation models include three types of failure rates: (a) instantaneous (all containers in a control volume fail at once), (b) uniformly distributed failures (containers fail at a linear rate between a specified starting and ending time), and (c) gaussian failure rates (containers fail at a rate determined by a mean failure time, standard deviation and gaussian distribution). Wasteform release models include four release mechanisms: (a) rinse with partitioning (inventory is released instantly upon container failure subject to equilibrium partitioning (sorption) with

  4. DWPF waste glass Product Composition Control System

    Energy Technology Data Exchange (ETDEWEB)

    Brown, K.G.; Postles, R.L.

    1992-07-01

    The Defense Waste Processing Facility (DWPF) will be used to blend aqueous radwaste (PHA) with solid radwaste (Sludge) in a waste receipt vessel (the SRAT). The resulting SRAT material is transferred to the SME an there blended with ground glass (Frit) to produce a batch of melter feed slurry. The SME material is passed to a hold tank (the MFT) which is used to continuously feed the DWPF melter. The melter. The melter produces a molten glass wasteform which is poured into stainless steel canisters for cooling and, ultimately, shipment to and storage in a geologic repository. The Product Composition Control System (PCCS) is the system intended to ensure that the melt will be processible and that the glass wasteform will be acceptable. This document provides a description of this system.

  5. DWPF waste glass Product Composition Control System

    Energy Technology Data Exchange (ETDEWEB)

    Brown, K.G.; Postles, R.L.

    1992-01-01

    The Defense Waste Processing Facility (DWPF) will be used to blend aqueous radwaste (PHA) with solid radwaste (Sludge) in a waste receipt vessel (the SRAT). The resulting SRAT material is transferred to the SME an there blended with ground glass (Frit) to produce a batch of melter feed slurry. The SME material is passed to a hold tank (the MFT) which is used to continuously feed the DWPF melter. The melter. The melter produces a molten glass wasteform which is poured into stainless steel canisters for cooling and, ultimately, shipment to and storage in a geologic repository. The Product Composition Control System (PCCS) is the system intended to ensure that the melt will be processible and that the glass wasteform will be acceptable. This document provides a description of this system.

  6. 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.

  7. Hazardous wastes in aquatic environments: Biological uptake and metabolism studies

    Energy Technology Data Exchange (ETDEWEB)

    Barber, J.; Apblett, A.; Ensley, H. [and others

    1996-05-02

    The projects discussed in this article include the following: the uptake, accumulation, metabolism, toxicity and physiological effects of various environmentally-important contaminants, inorganic and organic, in several wetland species that are interrelated through food webs; and investigation of the potential for developing and linking chemical and biological methods of remediation so as to encapsulate bioaccummulated ions in stable wasteforms such as ceramics and/or zeolites. 24 refs.

  8. OPC Paste Samples Exposed To Aggressive Solutions. Cementitious Barriers Partnership

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-11-01

    The study presented in this report focused on a low-activity wasteform containing a high pH pore solution with a significant level of sulfate. The purpose of the study was to improve understanding of the complex concrete/wasteform reactive transport problem, in particular the role of pH in sulfate attack. Paste samples prepared at three different water-to-cement ratios were tested. The mixtures were prepared with ASTM Type I cement, without additional admixtures. The samples were exposed to two different sodium sulfate contact solutions. The first solution was prepared at 0.15M Na2SO4. The second solution also incorporated 0.5M NaOH, to mimic the high pH conditions found in Saltstone. The data collected indicated that in Na2SO4 solution, damage occurs to the pastes. In the case of the high pH sulfate solution (Na2SO4 + NaOH), no signs of damage was observed on any of the paste mixtures. These results indicate that the high sulfate content found in the wasteform pore solution will not necessarily lead to severe damage to concrete. Good-quality mixtures could thus prove durable over the long term, and act as an effective barrier to prevent radionuclides from reaching the environment.

  9. OPC Paste Samples Exposed To Aggressive Solutions. Cementitious Barriers Partnership

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-11-01

    The study presented in this report focused on a low-activity wasteform containing a high-pH pore solution with a significant level of sulfate. The purpose of the study was to improve understanding of the complex concrete/wasteform reactive transport problem, in particular, the role of pH in sulfate attack. Paste samples prepared at three different water-to-cement ratios were tested. The mixtures were prepared with ASTM Type I cement, without additional admixtures. The samples were exposed to two different sodium sulfate contact solutions. The first solution was prepared at 0.15M Na2SO4. The second solution also incorporated 0.5M NaOH, to mimic the high pH conditions found in Saltstone. The data collected indicated that, in Na2SO4 solution, damage occurs to the pastes. In the case of the high-pH sulfate solution (Na2SO4 + NaOH), no signs of damage were observed on any of the paste mixtures. These results indicate that the high sulfate content found in the wasteform pore solution will not necessarily lead to severe damage to concrete. Good-quality mixtures could thus prove durable over the long term, and act as an effective barrier to prevent radionuclides from reaching the environment.

  10. Effect of pH on the release of radionuclides and chelating agents from cement-solidified decontamination ion-exchange resins collected from operating nuclear power stations

    Energy Technology Data Exchange (ETDEWEB)

    McIsaac, C.V.; Akers, D.W.; McConnell, J.W. (EG and G Idaho, Inc., Idaho Falls, ID (USA))

    1991-06-01

    Data are presented on the physical stability and leachability of radionuclides and chelating agents from cement-solidified decontamination ion-exchange resin wastes collected from two operating commercial light water reactors. Small-scale waste--form specimens collected during solidifications performed at the Brunswick Steam Electric Plant Unit 1 and at the James A. FitzPatrick Nuclear Power Station were leach-tested and subjected to compressive strength testing in accordance with the Nuclear Regulatory Commission's Technical Position on Waste Form'' (Revision 1). Samples of untreated resin waste collected from each solidification vessel before the solidification process were analyzed for concentrations of radionuclides, selected transition metals, and chelating agents to determine the quantities of these chemicals in the waste-form specimens. The chelating agents included oxalic, citric, and picolinic acids. In order to determine the effect of leachant chemical composition and pH on the stability and leachability of the waste forms, waste-form specimens were leached in various leachants. Results of this study indicate that differences in pH do not affect releases from cement-solidified decontamination ion-exchange resin waste forms, but that differences in leachant chemistry and the presence of chelating agents may affect the releases of radionuclides and chelating agents. Also, this study indicates that the cumulative releases of radionuclides and chelating agents are similar for waste- form specimens that decomposed and those that retained their general physical form. 36 refs., 60 figs., 28 tabs.

  11. Evaluation of final waste forms and recommendations for baseline alternatives to group and glass

    Energy Technology Data Exchange (ETDEWEB)

    Bleier, A.

    1997-09-01

    An assessment of final waste forms was made as part of the Federal Facilities Compliance Agreement/Development, Demonstration, Testing, and Evaluation (FFCA/DDT&E) Program because supplemental waste-form technologies are needed for the hazardous, radioactive, and mixed wastes of concern to the Department of Energy and the problematic wastes on the Oak Ridge Reservation. The principal objective was to identify a primary waste-form candidate as an alternative to grout (cement) and glass. The effort principally comprised a literature search, the goal of which was to establish a knowledge base regarding four areas: (1) the waste-form technologies based on grout and glass, (2) candidate alternatives, (3) the wastes that need to be immobilized, and (4) the technical and regulatory constraints on the waste-from technologies. This report serves, in part, to meet this goal. Six families of materials emerged as relevant; inorganic, organic, vitrified, devitrified, ceramic, and metallic matrices. Multiple members of each family were assessed, emphasizing the materials-oriented factors and accounting for the fact that the two most prevalent types of wastes for the FFCA/DDT&E Program are aqueous liquids and inorganic sludges and solids. Presently, no individual matrix is sufficiently developed to permit its immediate implementation as a baseline alternative. Three thermoplastic materials, sulfur-polymer cement (inorganic), bitumen (organic), and polyethylene (organic), are the most technologically developed candidates. Each warrants further study, emphasizing the engineering and economic factors, but each also has limitations that regulate it to a status of short-term alternative. The crystallinity and flexible processing of sulfur provide sulfur-polymer cement with the highest potential for short-term success via encapsulation. Long-term immobilization demands chemical stabilization, which the thermoplastic matrices do not offer. Among the properties of the remaining

  12. Molecular environmental science using synchrotron radiation:Chemistry and physics of waste form materials

    Energy Technology Data Exchange (ETDEWEB)

    Lindle, Dennis W.; Shuh, David K.

    2005-02-28

    Production of defense-related nuclear materials has generated large volumes of complex chemical wastes containing a mixture of radionuclides. The disposition of these wastes requires conversion of the liquid and solid-phase components into durable, solid forms suitable for long-term immobilization [1]. Specially formulated glass compositions, many of which have been derived from glass developed for commercial purposes, and ceramics such as pyrochlores and apatites, will be the main recipients for these wastes. The performance characteristics of waste-form glasses and ceramics are largely determined by the loading capacity for the waste constituents (radioactive and non-radioactive) and the resultant chemical and radiation resistance of the waste-form package to leaching (durability). There are unique opportunities for the use of near-edge soft-x-ray absorption fine structure (NEXAFS) spectroscopy to investigate speciation of low-Z elements forming the backbone of waste-form glasses and ceramics. Although nuclear magnetic resonance (NMR) is the primary technique employed to obtain speciation information from low-Z elements in waste forms, NMR is incompatible with the metallic impurities contained in real waste and is thus limited to studies of idealized model systems. In contrast, NEXAFS can yield element-specific speciation information from glass constituents without sensitivity to paramagnetic species. Development and use of NEXAFS for eventual studies of real waste glasses has significant implications, especially for the low-Z elements comprising glass matrices [5-7]. The NEXAFS measurements were performed at Beamline 6.3.1, an entrance-slitless bend-magnet beamline operating from 200 eV to 2000 eV with a Hettrick-Underwood varied-line-space (VLS) grating monochromator, of the Advanced Light Source (ALS) at LBNL. Complete characterization and optimization of this beamline was conducted to enable high-performance measurements.

  13. Leaching Rate Test of Nuclei Cs and Sr

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    A novel material-resemble verifiable cement for disposal of spent nuclear fuel reprocessing mediate-level waste has been developed. Waste form has been characterized for their compressive strength, phase composition. The cement formulation has been patented. In this experiment, the cement was mixed with simulated wastes for each composition 5 min at least. Ratio of waste to the cement is 0.45-0.55. After being packed into cylindrical molds, the grouts were cured for a period 28 days in a room temperature curing chamber at atmospheric pressure. The wasteform then

  14. Conditions and processes affecting radionuclide transport

    Science.gov (United States)

    Simmons, Ardyth M.; Neymark, Leonid A.

    2012-01-01

    Characteristics of host rocks, secondary minerals, and fluids would affect the transport of radionuclides from a previously proposed repository at Yucca Mountain, Nevada. Minerals in the Yucca Mountain tuffs that are important for retarding radionuclides include clinoptilolite and mordenite (zeolites), clay minerals, and iron and manganese oxides and hydroxides. Water compositions along flow paths beneath Yucca Mountain are controlled by dissolution reactions, silica and calcite precipitation, and ion-exchange reactions. Radionuclide concentrations along flow paths from a repository could be limited by (1) low waste-form dissolution rates, (2) low radionuclide solubility, and (3) radionuclide sorption onto geological media.

  15. Development of Advanced Electrochemical Emission Spectroscopy for Monitoring Corrosion in Simulated DOE Liquid Waste

    Energy Technology Data Exchange (ETDEWEB)

    MacDonal, Digby D.; Marx, Brian M.; Ahn, Sejin; Ruiz, Julio de; Soundararajan, Balaji; Smith, Morgan; Coulson, Wendy

    2005-06-15

    Various forms of general and localized corrosion represent principal threats to the integrity of DOE liquid waste storage tanks. These tanks, which are of a single wall or double wall design, depending upon their age, are fabricated from welded carbon steel and contain a complex waste-form comprised of NaOH and NaNO3, along with trace amounts of phosphate, sulfate, carbonate, and chloride. Because waste leakage can have a profound environmental impact, considerable interest exists in predicting the accumulation of corrosion damage, so as to more effectively schedule maintenance and repair.

  16. Radiological performance assessment for the E-Area Vaults Disposal Facility

    Energy Technology Data Exchange (ETDEWEB)

    Cook, J.R.

    2000-04-11

    This report is the first revision to ``Radiological Performance Assessment for the E-Area Vaults Disposal Facility, Revision 0'', which was issued in April 1994 and received conditional DOE approval in September 1994. The title of this report has been changed to conform to the current name of the facility. The revision incorporates improved groundwater modeling methodology, which includes a large data base of site specific geotechnical data, and special Analyses on disposal of cement-based wasteforms and naval wastes, issued after publication of Revision 0.

  17. Compatibility of technologies with regulations in the waste management of H-3, I-129, C-14, and Kr-85. Part I. Initial information base

    Energy Technology Data Exchange (ETDEWEB)

    Trevorrow, L.E.; Vandegrift, G.F.; Kolba, V.M.; Steindler, M.J.

    1983-08-01

    This report summarizes the information base that was collected and reviewed in preparation for carrying out an analysis of the compatibility with regulations of waste management technologies for disposal of H-3, I-129, C-14, and Kr-85. Based on the review of this literature, summaries are presented here of waste-form characteristics, packaging, transportation, and disposal methods. Also discussed are regulations that might apply to all operations involved in disposal of the four nuclides, including the processing of irradiated fuel in a fuel reprocessing plant, packaging, storage, transport, and final disposal. The compliance assessment derived from this information is reported in a separate document. 309 references.

  18. Building flexibility into the design of a pilot plant for the immobilisation of Pu containing residues and wastes

    Energy Technology Data Exchange (ETDEWEB)

    Scales, C R; Maddrell, E R [NNL, Havelock Rd, Workington, CA14 3YQ (United Kingdom); Hobbs, J; Stephen, R [Sellafield Ltd, Sellafield, CA20 1 PG (United Kingdom); Moricca, S; Stewart, M W A [ANSTO, New Illawara Road, Lucas Heights 2234 (Australia)

    2013-07-01

    NNL and ANSTO on behalf of Sellafield Ltd have developed a process for the immobilisation of a range of Pu containing wastes and residues. Following the inactive demonstration of the technology the project is now focusing on the design of an active pilot plant capable of validating the technology and ultimately immobilising a waste inventory containing around 100 kg plutonium. The diverse wastes from which it is uneconomic to recover Pu, require a flexible process with a wide product envelope capable of producing a wasteform suitable for disposal in a UK repository. Ceramics, glass ceramics and metal encapsulated waste-forms can be delivered by the process line which incorporates size reduction and heat treatment techniques with the aim of feeding a hot isostatic pressing process designed to deliver the highly durable waste-forms. Following a demonstration of feasibility, flowsheet development is progressing to support the design which has the aim of a fully flexible facility based in NNL's Central Laboratory on the Sellafield site. Optimisation of the size reduction, mixing and blending operations is being carried out using UO{sub 2} as a surrogate for PuO{sub 2}. This work is supporting the potential of using an enhanced glass ceramic formulation in place of the full ceramic with the aim of simplifying glove box operations. Heat treatment and subsequent HIPing strategies are being explored in order to eliminate any carbon from the feeds without increasing the valence state of the uranium present in some of the inventory which can result in an unwanted increase in wasteform volumes. The HIP and ancillary systems are being specifically designed to meet the requirements of the Sellafield site and within the constraints of the NNL Central Laboratory. The HIP is being configured to produce consolidated product cans consistent with the requirements of ongoing storage and disposal. With the aim of one cycle per day, the facility will deliver its mission of

  19. Actinide speciation in glass leach-layers: An EXAFS study

    Energy Technology Data Exchange (ETDEWEB)

    Biwer, B.M.; Soderholm, L. [Argonne National Lab., IL (United States); Greegor, R.B. [Boeing Co., Seattle, WA (United States); Lytle, F.W. [EXAFS Co., Pioche, NV (United States)

    1996-12-31

    Uranium L{sub 3} X-ray absorption data were obtained from two borosilicate glasses, which are considered as models for radioactive wasteforms, both before and after leaching. Surface sensitivity to uranium speciation was attained by a novel application of simultaneous fluorescence and electron-yield detection. Changes in speciation are clearly discernible, from U(VI) in the bulk to (UO{sub 2}){sup 2+}-uranyl in the leach layer. The leach-layer uranium concentration variations with leaching times are also determined from the data.

  20. Bases, Assumptions, and Results of the Flowsheet Calculations for the Decision Phase Salt Disposition Alternatives

    Energy Technology Data Exchange (ETDEWEB)

    Dimenna, R.A.; Jacobs, R.A.; Taylor, G.A.; Durate, O.E.; Paul, P.K.; Elder, H.H.; Pike, J.A.; Fowler, J.R.; Rutland, P.L.; Gregory, M.V.; Smith III, F.G.; Hang, T.; Subosits, S.G.; Campbell, S.G.

    2001-03-26

    The High Level Waste (HLW) Salt Disposition Systems Engineering Team was formed on March 13, 1998, and chartered to identify options, evaluate alternatives, and recommend a selected alternative(s) for processing HLW salt to a permitted wasteform. This requirement arises because the existing In-Tank Precipitation process at the Savannah River Site, as currently configured, cannot simultaneously meet the HLW production and Authorization Basis safety requirements. This engineering study was performed in four phases. This document provides the technical bases, assumptions, and results of this engineering study.

  1. Selection of models to calculate the LLW source term

    Energy Technology Data Exchange (ETDEWEB)

    Sullivan, T.M. (Brookhaven National Lab., Upton, NY (United States))

    1991-10-01

    Performance assessment of a LLW disposal facility begins with an estimation of the rate at which radionuclides migrate out of the facility (i.e., the source term). The focus of this work is to develop a methodology for calculating the source term. In general, the source term is influenced by the radionuclide inventory, the wasteforms and containers used to dispose of the inventory, and the physical processes that lead to release from the facility (fluid flow, container degradation, wasteform leaching, and radionuclide transport). In turn, many of these physical processes are influenced by the design of the disposal facility (e.g., infiltration of water). The complexity of the problem and the absence of appropriate data prevent development of an entirely mechanistic representation of radionuclide release from a disposal facility. Typically, a number of assumptions, based on knowledge of the disposal system, are used to simplify the problem. This document provides a brief overview of disposal practices and reviews existing source term models as background for selecting appropriate models for estimating the source term. The selection rationale and the mathematical details of the models are presented. Finally, guidance is presented for combining the inventory data with appropriate mechanisms describing release from the disposal facility. 44 refs., 6 figs., 1 tab.

  2. A literature review of mixed waste components: Sensitivities and effects upon solidification/stabilization in cement-based matrices

    Energy Technology Data Exchange (ETDEWEB)

    Mattus, C.H.; Gilliam, T.M.

    1994-03-01

    The US DOE Oak Ridge Field Office has signed a Federal Facility Compliance Agreement (FFCA) regarding Oak Ridge Reservation (ORR) mixed wastes subject to the land disposal restriction (LDR) provisions of the Resource conservation and Recovery Act. The LDR FFCA establishes an aggressive schedule for conducting treatability studies and developing treatment methods for those ORR mixed (radioactive and hazardous) wastes listed in Appendix B to the Agreement. A development, demonstration, testing, and evaluation program has been initiated to provide those efforts necessary to identify treatment methods for all of the wastes that meet Appendix B criteria. The program has assembled project teams to address treatment development needs in a variety of areas, including that of final waste forms (i.e., stabilization/solidification processes). A literature research has been performed, with the objective of determining waste characterization needs to support cement-based waste-form development. The goal was to determine which waste species are problematic in terms of consistent production of an acceptable cement-based waste form and at what concentrations these species become intolerable. The report discusses the following: hydration mechanisms of Portland cement; mechanisms of retardation and acceleration of cement set-factors affecting the durability of waste forms; regulatory limits as they apply to mixed wastes; review of inorganic species that interfere with the development of cement-based waste forms; review of radioactive species that can be immobilized in cement-based waste forms; and review of organic species that may interfere with various waste-form properties.

  3. Evolution in performance assessment modeling as a result of regulatory review

    Energy Technology Data Exchange (ETDEWEB)

    Rowat, J.H.; Dolinar, G.M.; Stephens, M.E. [AECL Chalk River Labs., Ontario (Canada)] [and others

    1995-12-31

    AECL is planning to build the IRUS (Intrusion Resistant Underground Structure) facility for near-surface disposal of LLRW. The PSAR (preliminary safety assessment report) was subject to an initial regulatory review during mid-1992. The regulatory authority provided comments on many aspects of the safety assessment documentation including a number of questions on specific PA (Performance Assessment) modelling assumptions. As a result of these comments as well as a separate detailed review of the IRUS disposal concept, changes were made to the conceptual and mathematical models. The original disposal concept included a non-sorbing vault backfill, with a strong reliance on the wasteform as a barrier. This concept was altered to decrease reliance on the wasteform by replacing the original backfill with a sand/clinoptilolite mix, which is a better sorber of metal cations. This change lead to changes in the PA models which in turn altered the safety case for the facility. This, and other changes that impacted performance assessment modelling are the subject of this paper.

  4. Incorporation of thorium in the rhabdophane structure: Synthesis and characterization of Pr1-2xCaxThxPO4·nH2O solid solutions

    Science.gov (United States)

    Qin, Danwen; Mesbah, Adel; Gausse, Clémence; Szenknect, Stéphanie; Dacheux, Nicolas; Clavier, Nicolas

    2017-08-01

    Thorium incorporation in the rhabdophane structure as Pr1-2xCaxThxPO4·nH2O solid solutions was successfully achieved and resulted in the preparation of a low temperature precursor of the monazite-cheralite type Pr1-2xCaxThxPO4. The rhabdophane compounds are considered as potential neoformed phases in case of release of actinides from the phosphate-based ceramic wasteforms envisaged to host radionuclides in the back-end of the nuclear fuel cycle. A multiparametric study was thus undertaken to specify the wet chemistry conditions (starting stoichiometry, temperature, heating time) leading to single phase Pr1-2xCaxThxPO4·nH2O powdered samples. The excess of calcium appeared to be a prevailing factor with a suggested initial Ca:Th ratio being equal to 10. Similarly, the recommended heating time should exceed 4 days while the optimal temperature of synthesis is 110 °C. Under these conditions, the stability domain of Pr1-2xCaxThxPO4·nH2O ranged from x = 0.00 to x = 0.15. After heating at 1100 °C under air during 6 h, rhabdophane-type samples were fully converted into the highly durable Pr1-2xCaxThxPO4 cheralite ceramic wasteform.

  5. Oak Ridge National Laboratory West End Treatment Facility simulated sludge vitrification demonstration, Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Cicero, C.A.; Bickford, D.F. [Westinghouse Savannah River Co., Aiken, SC (United States); Bennert, D.M.; Overcamp, T.J. [Clemson Univ., Anderson, SC (United States). Dept. of Environmental Systems Engineering

    1994-01-26

    Technologies are being developed by the US Department of Energy`s (DOE) Nuclear Facility sites to convert hazardous and mixed wastes to a form suitable for permanent disposal. Vitrification, which has been declared the Best Demonstrated Available Technology for high-level radioactive waste disposal by the EPA, is capable of producing a highly durable wasteform that minimizes disposal volumes through organic destruction, moisture evaporation, and porosity reduction. However, this technology must be demonstrated over a range of waste characteristics, including compositions, chemistries, moistures, and physical characteristics to ensure that it is suitable for hazardous and mixed waste treatment. These wastes are typically wastewater treatment sludges that are categorized as listed wastes due to the process origin or organic solvent content, and usually contain only small amounts of hazardous constituents. The Oak Ridge National Laboratory`s (ORNL) West End Treatment Facility`s (WETF) sludge is considered on of these representative wastes. The WETF is a liquid waste processing plant that generates sludge from the biodenitrification and precipitation processes. An alternative wasteform is needed since the waste is currently stored in epoxy coated carbon steel tanks, which have a limited life. Since this waste has characteristics that make it suitable for vitrification with a high likelihood of success, it was identified as a suitable candidate by the Mixed Waste Integrated Program (MWIP) for testing at CU. The areas of special interest with this sludge are (1) minimum nitrates, (2) organic destruction, and (3) waste water treatment sludges containing little or no filter aid.

  6. Disposal Unit Source Term (DUST) data input guide

    Energy Technology Data Exchange (ETDEWEB)

    Sullivan, T.M. [Brookhaven National Lab., Upton, NY (United States)

    1993-05-01

    Performance assessment of a low-level waste (LLW) disposal facility begins with an estimation of the rate at which radionuclides migrate out of the facility (i.e., the source term). The focus of this work is to develop a methodology for calculating the source term. In general, the source term is influenced by the radionuclide inventory, the wasteforms and containers used to dispose of the inventory, and the physical processes that lead to release from the facility (fluid flow, container degradation, wasteform leaching, and radionuclide transport). The computer code DUST (Disposal Unit Source Term) has been developed to model these processes. This document presents the models used to calculate release from a disposal facility, verification of the model, and instructions on the use of the DUST code. In addition to DUST, a preprocessor, DUSTIN, which helps the code user create input decks for DUST and a post-processor, GRAFXT, which takes selected output files and plots them on the computer terminal have been written. Use of these codes is also described.

  7. Compression and immersion tests and leaching of radionuclides, stable metals, and chelating agents from cement-solidified decontamination waste collected from nuclear power stations

    Energy Technology Data Exchange (ETDEWEB)

    Akers, D.W.; Kraft, N.C.; Mandler, J.W. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

    1994-06-01

    A study was performed for the Nuclear Regulatory Commission (NRC) to evaluate structural stability and leachability of radionuclides, stable metals, and chelating agents from cement-solidified decontamination ion-exchange resin wastes collected from seven commercial boiling water reactors and one pressurized water reactor. The decontamination methods used at the reactors were the Can-Decon, AP/Citrox, Dow NS-1, and LOMI processes. Samples of untreated resin waste and solidified waste forms were subjected to immersion and compressive strength testing. Some waste-form samples were leach-tested using simulated groundwaters and simulated seawater for comparison with the deionized water tests that are normally performed to assess waste-form leachability. This report presents the results of these tests and assesses the effects of the various decontamination methods, waste form formulations, leachant chemical compositions, and pH of the leachant on the structural stability and leachability of the waste forms. Results indicate that releases from intact and degraded waste forms are similar and that the behavior of some radionuclides such as {sup 55}Fe, {sup 60}Co, and {sup 99}Tc were similar. In addition, the leachability indexes are greater than 6.0, which meets the requirement in the NRC`s ``Technical Position on Waste Form,`` Revision 1.

  8. Migration of ions in cement paste as studied by SIMS

    Energy Technology Data Exchange (ETDEWEB)

    Prince, K.E.; Aldridge, L.P. [Australian Nuclear Science and Technology Organisation (ANSTO), Lucas Heights, NSW (Australia); Rougeron, P. [Electricite de France Direction des Etudes et Recherches, Les Renardiers (France)

    1998-06-01

    Cement is often used to condition and encapsulate low level radioactive waste before it is disposed of in a repository. Ground water can attack these waste-forms by transporting aggressive ions into the cement paste and by removing radioactive ions from the paste. The extent of the attack will be governed by the diffusion of the ions in the cement paste. In this study we examine the migration of aggressive carbonate ions and inactive Cs and Sr through cement pastes. The use of SIMS for establishing the penetration depths and diffusion profiles for Cs and Sr in cement will be explored. The penetration profiles of Cs and Sr in a non-zeolite cement paste were examined and compared to those of a paste made with zeolite. The effects of the non-homogeneous nature of the cement was most pronounced in the study of the zeolite rich cement; Cs being preferentially accumulated in the zeolite material. (authors). 4 refs., 2 figs.

  9. Magnesium alloys and graphite wastes encapsulated in cementitious materials: Reduction of galvanic corrosion using alkali hydroxide activated blast furnace slag.

    Science.gov (United States)

    Chartier, D; Muzeau, B; Stefan, L; Sanchez-Canet, J; Monguillon, C

    2017-03-15

    Magnesium alloys and graphite from spent nuclear fuel have been stored together in La Hague plant. The packaging of these wastes is under consideration. These wastes could be mixed in a grout composed of industrially available cement (Portland, calcium aluminate…). Within the alkaline pore solution of these matrixes, magnesium alloys are imperfectly protected by a layer of Brucite resulting in a slow corrosion releasing hydrogen. As the production of this gas must be considered for the storage safety, and the quality of wasteform, it is important to select a cement matrix capable of lowering the corrosion kinetics. Many types of calcium based cements have been tested and most of them have caused strong hydrogen production when magnesium alloys and graphite are conditioned together because of galvanic corrosion. Exceptions are binders based on alkali hydroxide activated ground granulated blast furnace slag (BFS) which are presented in this article. Copyright © 2016 Elsevier B.V. All rights reserved.

  10. Graphite immobilisation in iron phosphate glass composite materials produced by microwave and conventional sintering routes

    Energy Technology Data Exchange (ETDEWEB)

    Mayzan, M.Z.H. [Immobilisation Science Laboratory, Department of Materials Science and Engineering, The University of Sheffield, Sir Robert Hadfield Building, Mappin Street, Sheffield S1 3JD (United Kingdom); Faculty of Science, Technology and Human Development, Universiti Tun Hussein Onn Malaysia, 86400 Parit Raja, Batu Pahat, Johor (Malaysia); Stennett, M.C.; Hyatt, N.C. [Immobilisation Science Laboratory, Department of Materials Science and Engineering, The University of Sheffield, Sir Robert Hadfield Building, Mappin Street, Sheffield S1 3JD (United Kingdom); Hand, R.J., E-mail: r.hand@sheffield.ac.uk [Immobilisation Science Laboratory, Department of Materials Science and Engineering, The University of Sheffield, Sir Robert Hadfield Building, Mappin Street, Sheffield S1 3JD (United Kingdom)

    2014-11-15

    An investigation of microwave and conventional processing of iron phosphate based graphite glass composite materials as potential wasteforms for the immobilisation of irradiated graphite is reported. For the base iron phosphate glass, full reaction of the raw materials and formation of a glass melt occurs with consequent removal of porosity at 8 min microwave processing. When graphite is present, iron phosphate crystalline phases are formed with higher levels of residual porosity than in the sample prepared using conventional sintering under argon. It is found that graphite reacts with the microwave field when in powder form but this reaction is minimised when the graphite is incorporated into a pellet, and that the graphite also impedes sintering of the glass. Mössbauer spectroscopy indicates that reduction of iron also occurs with concomitant graphite oxidation. Conventionally sintered samples had lower porosities than the equivalent microwaved ones.

  11. Low Temperature Waste Immobilization Testing Vol. I

    Energy Technology Data Exchange (ETDEWEB)

    Russell, Renee L.; Schweiger, Michael J.; Westsik, Joseph H.; Hrma, Pavel R.; Smith, D. E.; Gallegos, Autumn B.; Telander, Monty R.; Pitman, Stan G.

    2006-09-14

    The Pacific Northwest National Laboratory (PNNL) is evaluating low-temperature technologies to immobilize mixed radioactive and hazardous waste. Three waste forms—alkali-aluminosilicate hydroceramic cement, “Ceramicrete” phosphate-bonded ceramic, and “DuraLith” alkali-aluminosilicate geopolymer—were selected through a competitive solicitation for fabrication and characterization of waste-form properties. The three contractors prepared their respective waste forms using simulants of a Hanford secondary waste and Idaho sodium bearing waste provided by PNNL and characterized their waste forms with respect to the Toxicity Characteristic Leaching Procedure (TCLP) and compressive strength. The contractors sent specimens to PNNL, and PNNL then conducted durability (American National Standards Institute/American Nuclear Society [ANSI/ANS] 16.1 Leachability Index [LI] and modified Product Consistency Test [PCT]) and compressive strength testing (both irradiated and as-received samples). This report presents the results of these characterization tests.

  12. Immobilization of fission products arising from pyrometallurgical reprocessing in chloride media

    Science.gov (United States)

    Leturcq, G.; Grandjean, A.; Rigaud, D.; Perouty, P.; Charlot, M.

    2005-12-01

    Spent nuclear fuel reprocessing to recover energy-producing elements such as uranium or plutonium can be performed by a pyrochemical process. In such method, the actinides and fission products are extracted by electrodeposition in a molten chloride medium. These processes generate chlorinated alkali salt flows contaminated by fission products, mainly Cs, Ba, Sr and rare earth elements constituting high-level waste. Two possible alternatives are investigated for managing this wasteform; a protocol is described for dechlorinating the fission products to allow vitrification, and mineral phases capable of immobilizing chlorides are listed to allow specification of a dedicated ceramic matrix suitable for containment of these chlorinated waste streams. The results of tests to synthesize chlorosilicate phases are also discussed.

  13. The Crystal Structure of Lanthanide Zirconates

    Science.gov (United States)

    Clements, Richard; Kennedy, Brendan; Ling, Christopher; Stampfl, Anton P. J.

    2010-03-01

    The lanthanide zirconates of composition Ln2Zr2O7 (Ln = La-Gd) are of interest for use in inert matrix fuels and nuclear wasteforms. The series undergoes a pyrochlore to fluorite phase transition as a function of the Ln atomic radii. The phase transition has been attributed to disordering of both the cation and the anion [1]. We have undertaken a synthesis of the lanthanide zirconate series Ln2Zr2O7 (Ln = La-Gd), Ln0.2Zr0.8O1.9 (Ln = Tb-Yb) and NdxHo2-xZr2O7 (0ANSTO's new high resolution powder diffractometer Echidna, in order to obtain accurate data on atomic displacement parameters and O 48f position across the series. These results will be presented, along with details of the analysis and synthetic techniques used.

  14. Glass Property Data and Models for Estimating High-Level Waste Glass Volume

    Energy Technology Data Exchange (ETDEWEB)

    Vienna, John D.; Fluegel, Alexander; Kim, Dong-Sang; Hrma, Pavel R.

    2009-10-05

    This report describes recent efforts to develop glass property models that can be used to help estimate the volume of high-level waste (HLW) glass that will result from vitrification of Hanford tank waste. The compositions of acceptable and processable HLW glasses need to be optimized to minimize the waste-form volume and, hence, to save cost. A database of properties and associated compositions for simulated waste glasses was collected for developing property-composition models. This database, although not comprehensive, represents a large fraction of data on waste-glass compositions and properties that were available at the time of this report. Glass property-composition models were fit to subsets of the database for several key glass properties. These models apply to a significantly broader composition space than those previously publised. These models should be considered for interim use in calculating properties of Hanford waste glasses.

  15. BLT-EC (Breach, Leach and Transport-Equilibrium Chemistry) data input guide. A computer model for simulating release and coupled geochemical transport of contaminants from a subsurface disposal facility

    Energy Technology Data Exchange (ETDEWEB)

    MacKinnon, R.J. [Brookhaven National Lab., Upton, NY (United States)]|[Ecodynamic Research Associates, Inc., Albuquerque, NM (United States); Sullivan, T.M.; Kinsey, R.R. [Brookhaven National Lab., Upton, NY (United States)

    1997-05-01

    The BLT-EC computer code has been developed, implemented, and tested. BLT-EC is a two-dimensional finite element computer code capable of simulating the time-dependent release and reactive transport of aqueous phase species in a subsurface soil system. BLT-EC contains models to simulate the processes (container degradation, waste-form performance, transport, chemical reactions, and radioactive production and decay) most relevant to estimating the release and transport of contaminants from a subsurface disposal system. Water flow is provided through tabular input or auxiliary files. Container degradation considers localized failure due to pitting corrosion and general failure due to uniform surface degradation processes. Waste-form performance considers release to be limited by one of four mechanisms: rinse with partitioning, diffusion, uniform surface degradation, and solubility. Transport considers the processes of advection, dispersion, diffusion, chemical reaction, radioactive production and decay, and sources (waste form releases). Chemical reactions accounted for include complexation, sorption, dissolution-precipitation, oxidation-reduction, and ion exchange. Radioactive production and decay in the waste form is simulated. To improve the usefulness of BLT-EC, a pre-processor, ECIN, which assists in the creation of chemistry input files, and a post-processor, BLTPLOT, which provides a visual display of the data have been developed. BLT-EC also includes an extensive database of thermodynamic data that is also accessible to ECIN. This document reviews the models implemented in BLT-EC and serves as a guide to creating input files and applying BLT-EC.

  16. Bubble Formation and Lattice Parameter Changes Resulting from He Irradiation of Defect-Fluorite Gd2Zr2O7

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, Caitlin A.; Patel, Maulik K.; Aguiar, Jeffery A.; Zhang, Yanwen; Crespillo, Miguel L.; Wen, Juan; Xue, Haizhou; Wang, Yongqiang; Weber, William J.

    2016-08-15

    Pyrochlores have long been considered as potential candidates for advanced ceramic waste-forms for the immobilization of radioactive waste nuclides. This work provides evidence that Gd2Zr2O7, often considered the most radiation tolerant pyrochlore, could be susceptible to radiation damage in the form of bubble nucleation at the highest He doses expected over geological time. Ion irradiations were utilized to experimentally simulate the radiation damage and He accumulation produced by ..alpha..-decay. Samples were pre-damaged using 7 MeV Au3+ to induce the pyrochlore to defect-fluorite phase transformation, which would occur due to ..alpha..-recoil damage within several hundred years of storage in a Gd2Zr2O7 waste-form. These samples were then implanted to various He concentrations in order to study the long-term effects of He accumulation. Helium bubbles 1-3 nm in diameter were observed in TEM at a concentration of 4.6 at.% He. Some bubbles remained isolated, while others formed chains 10-30 nm in length parallel to the surface. GIXRD measurements showed lattice swelling after irradiating pristine Gd2Zr2O7 with 7 MeV Au3+ to a fluence of 2.2 x 1015 Au/cm2. An increase in lattice swelling was also measured after 2.2 x 1015 Au/cm2 + 2 x 1015 He/cm2 and 2.2 x 1015 Au/cm2 + 2 x 1016 He/cm2. A decrease in lattice swelling was measured after irradiation with 2.2 x 1015 Au/cm2 + 2 x 1017 He/cm2, the fluence where bubbles and bubble chains were observed in TEM. Bubble chains are thought to form in order to reduce lattice strain normal to the surface, which is produced by the Au and He irradiation damage.

  17. Radioactive wastes dispersed in stabilized ash cements

    Energy Technology Data Exchange (ETDEWEB)

    Rubin, J.B.; Taylor, C.M.V.; Sivils, L.D.; Carey, J.W.

    1997-12-31

    One of the most widely-used methods for the solidification/stabilization of low-level radwaste is by incorporation into Type-I/II ordinary portland cement (OPC). Treating of OPC with supercritical fluid carbon dioxide (SCCO{sub 2}) has been shown to significantly increase the density, while simultaneously decreasing porosity. In addition, the process significantly reduces the hydrogenous content, reducing the likelihood of radiolytic decomposition reactions. This, in turn, permits increased actinide loadings with a concomitant reduction in disposable waste volume. In this article, the authors discuss the combined use of fly-ash-modified OPC and its treatment with SCCO{sub 2} to further enhance immobilization properties. They begin with a brief summary of current cement immobilization technology in order to delineate the areas of concern. Next, supercritical fluids are described, as they relate to these areas of concern. In the subsequent section, they present an outline of results on the application of SCCO{sub 2} to OPC, and its effectiveness in addressing these problem areas. Lastly, in the final section, they proffer their thoughts on why they believe, based on the OPC results, that the incorporation of fly ash into OPC, followed by supercritical fluid treatment, can produce highly efficient wasteforms.

  18. Taiwan industrial cooperation program technology transfer for low-level radioactive waste final disposal - phase I.

    Energy Technology Data Exchange (ETDEWEB)

    Knowlton, Robert G.; Cochran, John Russell; Arnold, Bill Walter; Jow, Hong-Nian; Mattie, Patrick D.; Schelling, Frank Joseph Jr. (; .)

    2007-01-01

    Sandia National Laboratories and the Institute of Nuclear Energy Research, Taiwan have collaborated in a technology transfer program related to low-level radioactive waste (LLW) disposal in Taiwan. Phase I of this program included regulatory analysis of LLW final disposal, development of LLW disposal performance assessment capabilities, and preliminary performance assessments of two potential disposal sites. Performance objectives were based on regulations in Taiwan and comparisons to those in the United States. Probabilistic performance assessment models were constructed based on limited site data using software including GoldSim, BLT-MS, FEHM, and HELP. These software codes provided the probabilistic framework, container degradation, waste-form leaching, groundwater flow, radionuclide transport, and cover infiltration simulation capabilities in the performance assessment. Preliminary performance assessment analyses were conducted for a near-surface disposal system and a mined cavern disposal system at two representative sites in Taiwan. Results of example calculations indicate peak simulated concentrations to a receptor within a few hundred years of LLW disposal, primarily from highly soluble, non-sorbing radionuclides.

  19. Grout for closure of the demonstration vault at the US DOE Hanford Facility. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Wakeley, L.D.; Ernzen, J.J.

    1992-08-01

    The Waterways Experiment Station (WES) developed a grout to be used as a cold- (nonradioactive) cap or void-fill grout between the solidified low-level waste and the cover blocks of a demonstration vault for disposal of phosphate-sulfate waste (PSW) at the US Department of Energy (DOE) Hanford Facility. The project consisted of formulation and evaluation of candidate grouts and selection of the best candidate grout, followed by a physical scale-model test to verify grout performance under project-specific conditions. Further, the project provided data to verify numerical models (accomplished elsewhere) of stresses and isotherms inside the Hanford demonstration vault. Evaluation of unhardened grout included obtaining data on segregation, bleeding, flow, and working time. For hardened grout, strength, volume stability, temperature rise, and chemical compatibility with surrogate wasteform grout were examined. The grout was formulated to accommodate unique environmental boundary conditions (vault temperature = 45 C) and exacting regulatory requirements (mandating less than 0.1% shrinkage with no expansion and no bleeding); and to remain pumpable for a minimum of 2 hr. A grout consisting of API Class H oil-well cement, an ASTM C 618 Class F fly ash, sodium bentonite clay, and a natural sand from the Hanford area met performance requirements in laboratory studies. It is recommended for use in the DOE Hanford demonstration PSW vault.

  20. Comparison of simulants to actual neutralized current acid waste: process and product testing of three NCAW core samples from Tanks 101-AZ and 102-AZ

    Energy Technology Data Exchange (ETDEWEB)

    Morrey, E.V.; Tingey, J.M.; Elliott, M.L.

    1996-10-01

    A vitrification plant is planned to process the high-level waste (HLW) solids from Hanford Site tanks into canistered glass logs for disposal in a national repository. Programs were established within the Pacific Northwest Laboratory Vitrification Technology Development (PVTD) Project to test and model simulated waste to support design, feed processability, operations, permitting, safety, and waste-form qualification. Parallel testing with actual radioactive waste was performed on a laboratory-scale to confirm the validity of using simulants and glass property models developed from simulants. Laboratory-scale testing has been completed on three radioactive core samples from tanks 101-AZ and 102-AZ containing neutralized current acid waste (NCAW), which is one of the first waste types to be processed in the high-level waste vitrification plant under a privatization scenario. Properties of the radioactive waste measured during process and product testing were compared to simulant properties and model predictions to confirm the validity of simulant and glass property ,models work. This report includes results from the three NCAW core samples, comparable results from slurry and glass simulants, and comparisons to glass property model predictions.

  1. Conasauga near-surface heater experiment. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Krumhansl, J.L.

    1979-11-01

    The Conasauga Experiment was undertaken to begin assessment of the thermomechanical and chemical response of a specific shale to the heat resulting from emplacement of high-level nuclear wastes. Canister-size heaters were implanted in Conasauga shale in Tennessee. Instrumentation arrays wee placed at various depths in drill holes around each heater. The heaters operated for 8 months and, after the first 4 days, were maintained at 385/sup 0/C. Emphasis was on characterizing the thermal and mechanical response of the formation. Conduction was the major mode of heat transport; convection was perceptible only at temperatures above the boiling point of water. Despite dehydration of the shale at higher temperatures, in situ thermal conductivity was essentially constant and not a function of temperature. The mechanical response of the formation was a slight overall expansion, apparently resulting in a general decrease in permeability. Metallurgical observations were made, the stability of a borosilicate glass wasteform simulant was assessed, and changes in formation mineralogy and groundwater composition were documented. In each of these areas, transient nonequilibrium processes occur that affect material stability and may be important in determining the integrity of a repository. In general, data from the test reflect favorably on the use of shale as a disposal medium for nuclear waste.

  2. Combined effects of radiation damage and He accumulation on bubble nucleation in Gd2Ti2O7

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, Caitlin A.; Patel, Maulik K.; Aguiar, Jeffery A.; Zhang, Yanwen; Crespillo, Miguel L.; Wen, Juan; Xue, Haizhou; Wang, Yongqiang; Weber, William J.

    2016-10-01

    Pyrochlores have long been considered as host phases for long-term immobilization of radioactive waste nuclides that would undergo ..alpha..-decay for hundreds of thousands of years. This work utilizes ion-beam irradiations to examine the combined effects of radiation damage and He accumulation on bubble formation in Gd2Ti2O7 over relevant waste-form timescales. Helium bubbles are not observed in pre-damaged Gd2Ti2O7 implanted with 2 x 1016 He/cm2, even after post-implantation irradiations with 7 MeV Au3+ at 300, 500, and 700 K. However, He bubbles with average diameters of 1.5 nm and 2.1 nm are observed in pre-damaged (amorphous) Gd2Ti2O7 and pristine Gd2Ti2O7, respectively, after implantation of 2 x 1017 He/cm2. The critical He concentration for bubble nucleation in Gd2Ti2O7 is estimated to be 6 at.% He.

  3. Comparison of simulants to actual neutralized current acid waste: Process and product testing of three NCAW core samples from Tanks 101-AZ and 102-AZ

    Energy Technology Data Exchange (ETDEWEB)

    Morrey, E.V.; Tingey, J.M.

    1996-04-01

    A vitrification plant is planned to process the high-level waste (HLW) solids from Hanford Site tanks into canistered glass logs for disposal in a national repository. Programs have been established within the Pacific Northwest Laboratory Vitrification Technology Development (PVTD) Project to test and model simulated waste to support design, feed processability, operations, permitting, safety, and waste-form qualification. Parallel testing with actual radioactive waste is being performed on a laboratory-scale to confirm the validity of using simulants and glass property models developed from simulants. Laboratory-scale testing has been completed on three radioactive core samples from tanks 101-AZ and 102-AZ containing neutralized current acid waste (NCAW), which is one of the first waste types to be processed in the high-level waste vitrification plant under a privatization scenario. Properties of the radioactive waste measured during process and product testing were compared to simulant properties and model predictions to confirm the validity of simulant and glass property models work. This report includes results from the three NCAW core samples, comparable results from slurry and glass simulants, and comparisons to glass property model predictions.

  4. Characterisation of Ba(OH){sub 2}–Na{sub 2}SO{sub 4}–blast furnace slag cement-like composites for the immobilisation of sulfate bearing nuclear wastes

    Energy Technology Data Exchange (ETDEWEB)

    Mobasher, Neda; Bernal, Susan A.; Hussain, Oday H. [Immobilisation Science Laboratory, Department of Materials Science and Engineering, The University of Sheffield, Sheffield S1 3JD (United Kingdom); Apperley, David C. [Solid-State NMR Group, Department of Chemistry, Durham University, Durham DH1 3LE (United Kingdom); Kinoshita, Hajime [Immobilisation Science Laboratory, Department of Materials Science and Engineering, The University of Sheffield, Sheffield S1 3JD (United Kingdom); Provis, John L., E-mail: j.provis@sheffield.ac.uk [Immobilisation Science Laboratory, Department of Materials Science and Engineering, The University of Sheffield, Sheffield S1 3JD (United Kingdom)

    2014-12-15

    Soluble sulfate ions in nuclear waste can have detrimental effects on cementitious wasteforms and disposal facilities based on Portland cement. As an alternative, Ba(OH){sub 2}–Na{sub 2}SO{sub 4}–blast furnace slag composites are studied for immobilisation of sulfate-bearing nuclear wastes. Calcium aluminosilicate hydrate (C–A–S–H) with some barium substitution is the main binder phase, with barium also present in the low solubility salts BaSO{sub 4} and BaCO{sub 3}, along with Ba-substituted calcium sulfoaluminate hydrates, and a hydrotalcite-type layered double hydroxide. This reaction product assemblage indicates that Ba(OH){sub 2} and Na{sub 2}SO{sub 4} act as alkaline activators and control the reaction of the slag in addition to forming insoluble BaSO{sub 4}, and this restricts sulfate availability for further reaction as long as sufficient Ba(OH){sub 2} is added. An increased content of Ba(OH){sub 2} promotes a higher degree of reaction, and the formation of a highly cross-linked C–A–S–H gel. These Ba(OH){sub 2}–Na{sub 2}SO{sub 4}–blast furnace slag composite binders could be effective in the immobilisation of sulfate-bearing nuclear wastes.

  5. Use of admixtures in organic-contaminated cement-clay pastes.

    Science.gov (United States)

    Gallo Stampino, Paola; Zampori, Luca; Dotelli, Giovanni; Meloni, Paola; Sora, Isabella Natali; Pelosato, Renato

    2009-01-30

    In this work microstructure, porosity and hydration degree of cement-based solidified/stabilized wasteforms were studied before assessing their leaching behaviour. 2-Chloroaniline was chosen as a model liquid organic pollutant and included into cement pastes, which were also modified with different admixtures for concrete: a superplasticizer based on acrylic-modified polymer, a synthetic rubber latex and a waterproofing agent. An organoclay, modified with an ammonium quaternary salt (benzyl-dimethyl-tallowammonium, BDMTA), was added to the pastes as pre-sorbent agent of the organic matter. All the samples were dried up to constant weight in order to stop the hydration process at different times during the first 28 days of curing, typically, after 1 day (1d), 7 days (7d) and 28 days. Then, the microstructure of the hardened cement-clay pastes was investigated by powder X-ray diffraction (XRD). The hydration degree and porosity were studied by thermal analysis (TG/DTA) and mercury intrusion porosimetry (MIP), respectively. For samples cured for 28 days a short-term leach test set by Italian regulation for industrial waste recycling (D.M. 5 February 1998) was performed. The best results showed a 5% release of the total initial amount of organic pollutant.

  6. Radioactive wastes dispersed in stabilized ash cements

    Energy Technology Data Exchange (ETDEWEB)

    Rubin, J.B.; Taylor, C.M.V.; Sivils, L.D.; Carey, J.W.

    1997-12-31

    One of the most widely-used methods for the solidification/stabilization of low-level radwaste is by incorporation into Type-I/II ordinary portland cement (OPC). Treating of OPC with supercritical fluid carbon dioxide (SCCO{sub 2}) has been shown to significantly increase the density, while simultaneously decreasing porosity. In addition, the process significantly reduces the hydrogenous content, reducing the likelihood of radiolytic decomposition reactions. This, in turn, permits increased actinide loadings with a concomitant reduction in disposable waste volume. In this article, the authors discuss the combined use of fly-ash-modified OPC and its treatment with SCCO{sub 2} to further enhance immobilization properties. They begin with a brief summary of current cement immobilization technology in order to delineate the areas of concern. Next, supercritical fluids are described, as they relate to these areas of concern. In the subsequent section, they present an outline of results on the application of SCCO{sub 2} to OPC, and its effectiveness in addressing these problem areas. Lastly, in the final section, they proffer their thoughts on why they believe, based on the OPC results, that the incorporation of fly ash into OPC, followed by supercritical fluid treatment, can produce highly efficient wasteforms.

  7. Chemical Engineering Division fuel cycle programs. Quarterly progress report, July-September 1979

    Energy Technology Data Exchange (ETDEWEB)

    Steindler, M.J.; Couture, R.A.; Flynn, K.F.; Jardine, L.J.; Mecham, W.J.; Pelto, R.H.; Seitz, M.G.; Williams, J.

    1980-09-01

    In a project to identify the advantages and disadvantages of encapsulating solidified waste forms in a metal matrix, leach rates of hazardous radionuclides from various matrix mterials as a function of temperature are being studied. Also, a methodology for analyzing particle size distributions obtained in impact-testing of brittle waste-form materials has been applied to the impact testing of Pyrex spheres and to earlier impact tests of a variety of materials. The transport properties of nuclear waste in geologic media are being studied. Porosity of basalt columns was measured by a method based on the elution of tritiated water. Batch tests were performed to determine the effect of rubidium concentration on cesium adsorption by limestone. An apparatus for infiltrating intact rocks with high-pressure groundwater solutions was constructed. In work on trace-element transport in lithic material, the sorption by Fe/sub 2/O/sub 3/ of iodate in concentrations of 10/sup -2/ to 10/sup -13/M and from pH 3 to 8.7 was measured, as was the sorption of iodate by sea sediments.

  8. Magnesium alloys and graphite wastes encapsulated in cementitious materials: Reduction of galvanic corrosion using alkali hydroxide activated blast furnace slag

    Energy Technology Data Exchange (ETDEWEB)

    Chartier, D., E-mail: david.chartier@cea.fr [Commissariat à l' Energie Atomique et aux Energies Alternatives, CEA, DEN, DTCD, SPDE, F-30207 Bagnols-sur-Cèze (France); Muzeau, B. [DEN-Service d’Etude du Comportement des Radionucléides (SECR), CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France); Stefan, L. [AREVA NC/D& S - France/Technical Department, 1 place Jean Millier 92084 Paris La Défense (France); Sanchez-Canet, J. [Commissariat à l' Energie Atomique et aux Energies Alternatives, CEA, DEN, DTCD, SPDE, F-30207 Bagnols-sur-Cèze (France); Monguillon, C. [DEN-Service d’Etude du Comportement des Radionucléides (SECR), CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette (France)

    2017-03-15

    Highlights: • Embedded in cement, magnesium is corroded by residual water present in porosity of the matrix. • Corrosion is enhanced by galvanic phenomenon when magnesium is in contact with graphite. • Galvanic corrosion of magnesium in contact with graphite debris is shown to be severe with ordinary Portland cement. • Galvanic corrosion is significantly lowered in high alkali medium such as sodium hydroxide. • Sodium hydroxide activated blast furnace slag is a convenient binder to embed magnesium. - Abstract: Magnesium alloys and graphite from spent nuclear fuel have been stored together in La Hague plant. The packaging of these wastes is under consideration. These wastes could be mixed in a grout composed of industrially available cement (Portland, calcium aluminate…). Within the alkaline pore solution of these matrixes, magnesium alloys are imperfectly protected by a layer of Brucite resulting in a slow corrosion releasing hydrogen. As the production of this gas must be considered for the storage safety, and the quality of wasteform, it is important to select a cement matrix capable of lowering the corrosion kinetics. Many types of calcium based cements have been tested and most of them have caused strong hydrogen production when magnesium alloys and graphite are conditioned together because of galvanic corrosion. Exceptions are binders based on alkali hydroxide activated ground granulated blast furnace slag (BFS) which are presented in this article.

  9. Grout to meet physical and chemical requirements for closure at Hanford grout vaults. Final report

    Energy Technology Data Exchange (ETDEWEB)

    1994-07-21

    The US Army Engineer Waterways Experiment Station (WES) developed a grout based on portland cement, Class F fly ash, and bentonite clay, for the Hanford Grout Vault Program. The purpose of this grout was to fill the void between a wasteform containing 106-AN waste and the vault cover blocks. Following a successful grout development program, heat output, volume change, and compressive strength were monitored with time in simulated repository conditions and in full-depth physical models. This research indicated that the cold-cap grout could achieve and maintain adequate volume stability and other required physical properties in the internal environment of a sealed vault. To determine if contact with 106-AN liquid waste would cause chemical deterioration of the cold-cap grout, cured specimens were immersed in simulated waste. Over a period of 21 days at 150 F, specimens increased in mass without significant changes in volume. X-ray diffraction of reacted specimens revealed crystallization of sodium aluminum silicate hydrate. Scanning electron microscopy used with X-ray fluorescence showed that clusters if this phase had formed in grout pores, increasing grout density and decreasing its effective porosity. Physical and chemical tests collectively indicate a sealing component. However, the Hanford Grout Vault Program was cancelled before completion of this research. This report summarizes close-out Waterways Experiment Station when the Program was cancelled.

  10. Final Report for the Demonstration of Plasma In-situ Vitrification at the 904-65G K-Reactor Seepage Basin

    Energy Technology Data Exchange (ETDEWEB)

    Blundy, R.F. [Westinghouse Savannah River Company, AIKEN, SC (United States); Zionkowki, P.G.

    1997-12-22

    The In-situ Vitrification (ISV) process potentially offers the most stable waste-form for containment of radiologically contaminated soils while minimizing personnel contamination. This is a problem that is extensive, and at the same time unique, to the US Department of Energy`s (DOE) Weapons Complex. An earlier ISV process utilized joule heating of the soil to generate the subsurface molten glass product. However previous test work has indicated that the Savannah river Site soils (SRS) may not be entirely suitable for vitrification by joule heating due to their highly refractory nature. The concept of utilizing a plasma torch for soil remediation by in-situ vitrification has recently been developed, and laboratory test work on a 100 kW unit has indicated a potentially successful application with SRS soils. The Environmental Restoration Division (ERD) of Westinghouse Savannah River Company (WSRC) conducted the first field scale demonstration of this process at the (904-65G) K-Reactor Seepage Basin in October 1996 with the intention of determining the applicability and economics of the process for remediation of a SRS radioactive seepage basin. The demonstration was successful in completing three vitrification runs, including two consecutive runs that fused together adjacent columns of glass to form a continuous monolith. This report describes the demonstration, documents the engineering data that was obtained, summarizes the process economics and makes recommendations for future development of the process and equipment.

  11. Interim Report on Development of a Model to Predict Dissolution Behavior of the Titanate Waste Form in a Repository

    Energy Technology Data Exchange (ETDEWEB)

    Bourcier, W.L.

    1999-08-16

    Dissolution testing performed to date on a titanate waste form under development for plutonium immobilization reveals the following: (1) The wasteform is very durable. Many of the test results have shown the dissolution rate to be below detection or less than background levels of the constituent elements; (2) elemental release is non-stoichiometric with Pu, U, Ca, and Gd released faster than Ti and Hf at most pH conditions; (3) dissolution rates measured in flow-through tests sometimes show a continuous decrease with time in tests of up to two years duration; (4) attempts to model the dissolution as a transport-controlled process with diffusion through a leached layer as the rate limiting mechanism show reasonable agreement at low pH conditions but poor agreement at neutral to alkaline pHs. Based on present uncertainties in our understanding of rate control, we have provided conservative estimates of radionuclide release rates based on the fastest observed release rates measured in short-term tests. These dissolution rates under repository-relevant conditions are in the range of 10{sup -3} to 10{sup -6}g/m{sup 2}/day.

  12. LITERATURE REVIEW: REDUCTION OF NP(V) TO NP (IV)-ALTERNATIVES TO FERROUS SULFAMATE

    Energy Technology Data Exchange (ETDEWEB)

    Kessinger, G.; Kyser, E.; Almond, P.

    2009-09-28

    The baseline approach to control of Np oxidation in UREX and PUREX separation processes is the reduction of Np(V) and Np(VI) to Np(IV) using ferrous sulfamate. Use of this reagent results in increased sulfur and iron concentrations in the liquid waste streams from the process. Presence of these two elements, especially sulfur, increases the complexity of the development of wasteforms for immobilizing these effluents. Investigations are underway to identify reductants that eliminate sulfur and iron from the Np reduction process. While there are a variety of chemical reductants that will reduce Np to Np(IV) in nitric acid media, the reaction rates for most are so slow that the reductants are not be feasible for use in an operating plant process. In an attempt to identify additional alternatives to ferrous sulfamate, a literature search and review was performed. Based on the results of the literature review, it is concluded that photochemical and catalytic processes should also be investigated to test the utility of these two approaches. The catalytic process could be investigated for use in conjunction with chemical oxidants to speed the reaction rates for reductants that react slowly, but would otherwise be appropriate replacements for ferrous sulfamate. The photochemical approach, which has received little attention during the past few decades, also shows promise, especially the photocatalytic approach that includes a catalyst, such as Pt supported on SiC, which can be used in tandem with an oxidant, for Np reduction.

  13. Engineered barrier environment, Yucca Mountain

    Energy Technology Data Exchange (ETDEWEB)

    Wilder, D.G. [Lawrence Livermore National Lab., CA (United States)

    1994-12-31

    The suitability of Yucca Mountain (YM) as a potential nuclear waste repository site will ultimately depend on how well it provides for isolation of the waste. Analysis of isolation capabilities of YM must consider interactions between natural and engineered systems. In addition, environmental conditions are important to EBS design, materials testing, selection, design criteria, and waste-form characterization. Studies of environmental interactions with the EBS, have emphasized processes and changed (not ambient) conditions resulting from interaction with waste, since these are the pertinent conditions for the EBS. The results of these studies indicate that the radioactive heat-of-decay from spent nuclear fuel will play a dominant role in the performance of a potential repository at Yucca Mountain. In addition, coupled hydrothermal-geochemical phenomena may significantly affect the performance of natural barriers surrounding the repository. Depending on the thermal-loading management strategy, as well as site conditions, repository heat may either substantially increase the likelihood of water contacting waste packages, with an associated potential increased magnitude of release and transport of radionuclides, or preclude, or at least minimize, these effects for extended periods of time, perhaps as much as hundreds of thousand years.

  14. Comparison of simulants to actual neutralized current acid waste: process and product testing of three NCAW core samples from Tanks 101-AZ and 102-AZ

    Energy Technology Data Exchange (ETDEWEB)

    Morrey, E.V.; Tingey, J.M.; Elliott, M.L.

    1996-10-01

    A vitrification plant is planned to process the high-level waste (HLW) solids from Hanford Site tanks into canistered glass logs for disposal in a national repository. Programs were established within the Pacific Northwest Laboratory Vitrification Technology Development (PVTD) Project to test and model simulated waste to support design, feed processability, operations, permitting, safety, and waste-form qualification. Parallel testing with actual radioactive waste was performed on a laboratory-scale to confirm the validity of using simulants and glass property models developed from simulants. Laboratory-scale testing has been completed on three radioactive core samples from tanks 101-AZ and 102-AZ containing neutralized current acid waste (NCAW), which is one of the first waste types to be processed in the high-level waste vitrification plant under a privatization scenario. Properties of the radioactive waste measured during process and product testing were compared to simulant properties and model predictions to confirm the validity of simulant and glass property ,models work. This report includes results from the three NCAW core samples, comparable results from slurry and glass simulants, and comparisons to glass property model predictions.

  15. Combined effects of radiation damage and He accumulation on bubble nucleation in Gd2Ti2O7

    Science.gov (United States)

    Taylor, Caitlin A.; Patel, Maulik K.; Aguiar, Jeffery A.; Zhang, Yanwen; Crespillo, Miguel L.; Wen, Juan; Xue, Haizhou; Wang, Yongqiang; Weber, William J.

    2016-10-01

    Pyrochlores have long been considered as host phases for long-term immobilization of radioactive waste nuclides that would undergo α-decay for hundreds of thousands of years. This work utilizes ion-beam irradiations to examine the combined effects of radiation damage and He accumulation on bubble formation in Gd2Ti2O7 over relevant waste-form timescales. Helium bubbles are not observed in pre-damaged Gd2Ti2O7 implanted with 2 × 1016 He/cm2, even after post-implantation irradiations with 7 MeV Au3+ at 300, 500, and 700 K. However, He bubbles with average diameters of 1.5 nm and 2.1 nm are observed in pre-damaged (amorphous) Gd2Ti2O7 and pristine Gd2Ti2O7, respectively, after implantation of 2 × 1017 He/cm2. The critical He concentration for bubble nucleation in Gd2Ti2O7 is estimated to be 6 at.% He.

  16. Memo, "Incorporation of HLW Glass Shell V2.0 into the Flowsheets," to ED Lee, CCN: 184905, October 20, 2009

    Energy Technology Data Exchange (ETDEWEB)

    Gimpel, Rodney F.; Kruger, Albert A.

    2013-12-18

    Efforts are being made to increase the efficiency and decrease the cost of vitrifying radioactive waste stored in tanks at the U.S. Department of Energy Hanford Site. The compositions of acceptable and processable high-level waste (HL W) glasses need to be optimized to minimize the waste-form volume and, hence, to reduce cost. A database of glass properties of waste glass and associated simulated waste glasses was collected and documented in PNNL 18501, Glass Property Data and Models for Estimating High-Level Waste Glass Volume and glass property models were curve-fitted to the glass compositions. A routine was developed that estimates HL W glass volumes using the following glass property models: II Nepheline, II One-Percent Crystal Temperature (T1%), II Viscosity (11) II Product Consistency Tests (PCT) for boron, sodium, and lithium, and II Liquidus Temperature (TL). The routine, commonly called the HL W Glass Shell, is presented in this document. In addition to the use of the glass property models, glass composition constraints and rules, as recommend in PNNL 18501 and in other documents (as referenced in this report) were incorporated. This new version of the HL W Glass Shell should generally estimate higher waste loading in the HL W glass than previous versions.

  17. VARIABILITY OF KD VALUES IN CEMENTITIOUS MATERIALS AND SEDIMENTS

    Energy Technology Data Exchange (ETDEWEB)

    Almond, P.; Kaplan, D.; Shine, E.

    2012-02-02

    Measured distribution coefficients (K{sub d} values) for environmental contaminants provide input data for performance assessments (PA) that evaluate physical and chemical phenomena for release of radionuclides from wasteforms, degradation of engineered components and subsequent transport of radionuclides through environmental media. Research efforts at SRNL to study the effects of formulation and curing variability on the physiochemical properties of the saltstone wasteform produced at the Saltstone Disposal Facility (SDF) are ongoing and provide information for the PA and Saltstone Operations. Furthermore, the range and distribution of plutonium K{sub d} values in soils is not known. Knowledge of these parameters is needed to provide guidance for stochastic modeling in the PA. Under the current SRS liquid waste processing system, supernate from F & H Tank Farm tanks is processed to remove actinides and fission products, resulting in a low-curie Decontaminated Salt Solution (DSS). At the Saltstone Production Facility (SPF), DSS is mixed with premix, comprised of blast furnace slag (BFS), Class F fly ash (FA), and portland cement (OPC) to form a grout mixture. The fresh grout is subsequently placed in SDF vaults where it cures through hydration reactions to produce saltstone, a hardened monolithic waste form. Variation in saltstone composition and cure conditions of grout can affect the saltstone's physiochemical properties. Variations in properties may originate from variables in DSS, premix, and water to premix ratio, grout mixing, placing, and curing conditions including time and temperature (Harbour et al. 2007; Harbour et al. 2009). There are no previous studies reported in the literature regarding the range and distribution of K{sub d} values in cementitious materials. Presently, the Savannah River Site (SRS) estimate ranges and distributions of K{sub d} values based on measurements of K{sub d} values made in sandy SRS sediments (Kaplan 2010). The actual

  18. Final environmental impact statement. Waste Isolation Pilot Plant

    Energy Technology Data Exchange (ETDEWEB)

    1980-10-01

    In accordance with the National Environmental Policy Act (NEPA) of 1969, the US Department of Energy (DOE) has prepared this document as environmental input to future decisions regarding the Waste Isolation Pilot Plant (WIPP), which would include the disposal of transuranic waste, as currently authorized. The alternatives covered in this document are the following: (1) Continue storing transuranic (TRU) waste at the Idaho National Engineering Laboratory (INEL) as it is now or with improved confinement. (2) Proceed with WIPP at the Los Medanos site in southeastern New Mexico, as currently authorized. (3) Dispose of TRU waste in the first available repository for high-level waste. The Los Medanos site would be investigated for its potential suitability as a candidate site. This is administration policy and is the alternative preferred by the DOE. (4) Delay the WIPP to allow other candidate sites to be evaluated for TRU-waste disposal. This environmental impact statement is arranged in the following manner: Chapter 1 is an overall summary of the analysis contained in the document. Chapters 2 and 4 set forth the objectives of the national waste-management program and analyze the full spectrum of reasonable alternatives for meeting these objectives, including the WIPP. Chapter 5 presents the interim waste-acceptance criteria and waste-form alternatives for the WIPP. Chapters 6 through 13 provide a detailed description and environmental analysis of the WIPP repository and its site. Chapter 14 describes the permits and approvals necessary for the WIPP and the interactions that have taken place with Federal, State, and local authorities, and with the general public in connection with the repository. Chapter 15 analyzes the many comments received on the DEIS and tells what has been done in this FEIS in response. The appendices contain data and discussions in support of the material in the text.

  19. Determination of organic products resulting of chemical and radiochemical decompositions of bitumen. Applications to embedded bitumens; Determination des produits organiques d'alterations chimiques et radiochimiques du bitume. Applications aux enrobes bitumes

    Energy Technology Data Exchange (ETDEWEB)

    Walczak, I

    2000-01-27

    Bitumen can be used for embedding most of wastes because of its high impermeability and its relatively low reactivity with of chemicals. Bituminization is one of selected solutions in agreement with nuclear safety, waste compatibility and economic criteria. Bitumen, during storage, undergoes an auto-irradiation due to embedded radio-elements. During this stage,drums are not airtight then oxygen is present. In disposal configuration, water, which is a potential vector of radioactivity and organic matter, is an other hazard factor liable to deteriorate the containment characteristics of bitumen wastes. The generation of water-soluble organic complexing agents can affect the integrity of the wasteform due to an increase of the radionuclides solubility. The first aim of this work is the quantitative and qualitative characterisation of soluble organic matter in bitumen leachates. Different leaching solutions were tested (various pH, ionic strength, ratio S/V). When the pH of the leaching solutions increases, the total organic carbon released increases as well. Identified molecules are aromatics like naphthalene, oxidised compounds like alcohols, linear carbonyls, aromatics, glycols and nitrogen compounds. For the cement equilibrated solution (pH 13.5), the effect of ionic strength becomes significative and influences the release of soluble organic matter. This soluble organic matter can be bio-degraded if microorganisms can growth. The second aim of this work is to study the effect of radio-oxidative ageing on the bitumen confinement properties. During radio-oxidation, the chemical properties of bitumen are modified. The {mu}-IRTF analysis shows the formation of hydroxyl compounds and aromatic acids. The formation of these polar groups does not influence in our study the water uptake. However the organic matter release increases significantly with the irradiation dose. (author)

  20. Verification of the integrity of barriers using gas diffusion

    Energy Technology Data Exchange (ETDEWEB)

    Ward, D.B. [SPECTRA Research Inst., Albuquerque, NM (United States); Williams, C.V. [Sandia National Labs., Albuquerque, NM (United States). Environmental Restoration Technologies Dept.

    1997-06-01

    In-situ barrier materials and designs are being developed for containment of high risk contamination as an alternative to immediate removal or remediation. The intent of these designs is to prevent the movement of contaminants in either the liquid or vapor phase by long-term containment, essentially buying time until the contaminant depletes naturally or a remediation can be implemented. The integrity of the resultant soil-binder mixture is typically assessed by a number of destructive laboratory tests (leaching, compressive strength, mechanical stability with respect to wetting and freeze-thaw cycles) which as a group are used to infer the likelihood of favorable long-term performance of the barrier. The need exists for a minimally intrusive yet quantifiable methods for assessment of a barrier`s integrity after emplacement, and monitoring of the barrier`s performance over its lifetime. Here, the authors evaluate non-destructive measurements of inert-gas diffusion (specifically, SF{sub 6}) as an indicator of waste-form integrity. The goals of this project are to show that diffusivity can be measured in core samples of soil jet-grouted with Portland cement, validate the experimental method through measurements on samples, and to calculate aqueous diffusivities from a series of diffusion measurements. This study shows that it is practical to measure SF{sub 6} diffusion rates in the laboratory on samples of grout (Portland cement and soil) typical of what might be used in a barrier. Diffusion of SF{sub 6} through grout (Portland cement and soil) is at least an order of magnitude slower than through air. The use of this tracer should be sensitive to the presence of fractures, voids, or other discontinuities in the grout/soil structure. Field-scale measurements should be practical on time-scales of a few days.

  1. Fundamental Science-Based Simulation of Nuclear Waste Forms

    Energy Technology Data Exchange (ETDEWEB)

    Devanathan, Ramaswami; Gao, Fei; Sun, Xin; Khaleel, Mohammad A.

    2010-10-04

    This report presents a hierarchical multiscale modeling scheme based on two-way information exchange. To account for all essential phenomena in waste forms over geological time scales, the models have to span length scales from nanometer to kilometer and time scales from picoseconds to millenia. A single model cannot cover this wide range and a multi-scale approach that integrates a number of different at-scale models is called for. The approach outlined here involves integration of quantum mechanical calculations, classical molecular dynamics simulations, kinetic Monte Carlo and phase field methods at the mesoscale, and continuum models. The ultimate aim is to provide science-based input in the form of constitutive equations to integrated codes. The atomistic component of this scheme is demonstrated in the promising waste form xenotime. Density functional theory calculations have yielded valuable information about defect formation energies. This data can be used to develop interatomic potentials for molecular dynamics simulations of radiation damage. Potentials developed in the present work show a good match for the equilibrium lattice constants, elastic constants and thermal expansion of xenotime. In novel waste forms, such as xenotime, a considerable amount of data needed to validate the models is not available. Integration of multiscale modeling with experimental work is essential to generate missing data needed to validate the modeling scheme and the individual models. Density functional theory can also be used to fill knowledge gaps. Key challenges lie in the areas of uncertainty quantification, verification and validation, which must be performed at each level of the multiscale model and across scales. The approach used to exchange information between different levels must also be rigorously validated. The outlook for multiscale modeling of wasteforms is quite promising.

  2. Dechlorination/Solidification of LiCl waste by using a synthetic inorganic composite with different compositions

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Na Young; Cho, In Hak; Park, Hwan Seo; Ahn, Do Hee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-09-15

    Waste salt generated from a pyro-processing for the recovery of uranium and transuranic elements has high volatility at vitrification temperature and low compatibility in conventional waste glasses. For this reason, KAERI (Korea Atomic Energy Research Institute) suggested a new method to de-chlorinate waste salt by using an inorganic composite named SAP (SiO{sub 2}-Al{sub 2}O{sub 3}-P{sub 2}O{sub 5}). In this study, the de-chlorination behavior of waste salt and the microstructure of consolidated form were examined by adding B{sub 2}O{sub 3} and Fe{sub 2}O{sub 3} to the original SAP composition. De-chlorination behavior of metal chloride waste was slightly changed with given compositions, compared with that of original SAP. In the consolidated forms, the phase separation between Si-rich phase and P-rich phase decreases with the amount of Al{sub 2}O{sub 3} or B{sub 2}O{sub 3} as a connecting agent between Si and P-rich phase. The results of PCT (Product Consistency Test) indicated that the leach-resistance of consolidated forms out of reference composition was lowered, even though the leach-resistance was higher than that of EA (Environmental Assessment) glass. From these results, it could be inferred that the change in the content of Al or B in U-SAP affected the microstructure and leach-resistance of consolidated form. Further studies related with correlation between composition and characteristics of wasteform are required for a better understanding.

  3. Release modes and processes relevant to source-term calculations at Yucca Mountain

    Energy Technology Data Exchange (ETDEWEB)

    Apted, M.J. [Intera Information Technologies, Denver (United States)

    1994-12-31

    The feasibility of permanent disposal of radioactive high-level waste (HLW) in repositories located in deep geologic formations is being studied world-wide. The most credible release pathway is interaction between groundwater and nuclear waste forms, followed by migration of radionuclide-bearing groundwater to the accessible environment. Under hydrologically unsaturated conditions, vapor transport of volatile radionuclides is also possible. The near-field encompasses the waste packages composed of engineered barriers, while the far-field includes the natural barriers. Taken together, these two subsystems define a series of multiple, redundant barriers that act to assure the safe isolation of nuclear waste. In the U.S., the Department of energy (DOE) is investigating the feasibility of safe, long-term disposal of high-level nuclear waste at the Yucca Mountain site in Nevada. The proposed repository horizon is located in non-welded tuffs within the unsaturated zone (i.e. above the water table) at Yucca Mountain. The purpose of this paper is to describe the source-term models for radionuclide release from waste packages at Yucca Mountain site. The first section describes the conceptual release modes that are relevant for this site and waste package design, based on a consideration of the performance of currently proposed engineered barriers under expected and unexpected conditions. No attempt is made to asses the reasonableness nor probability of occurrence for any specific release mode. The following section reviews the waste-form characteristics that are required to model and constrain the release of radionuclides from the waste package. The next section present mathematical models for the conceptual release modes, selected from those that have been implemented into a probabilistic total system assessment code developed for the Electric Power Research Institute (EPRI). (author) 4 figs., 35 refs.

  4. TECHNICAL ASSESSMENT OF BULK VITRIFICATION PROCESS & PRODUCT FOR TANK WASTE TREATMENT AT THE DEPARTMENT OF ENERGY HANFORD SITE

    Energy Technology Data Exchange (ETDEWEB)

    SCHAUS, P.S.

    2006-07-21

    At the U.S. Department of Energy (DOE) Hanford Site, the Waste Treatment Plant (WTP) is being constructed to immobilize both high-level waste (IUW) for disposal in a national repository and low-activity waste (LAW) for onsite, near-surface disposal. The schedule-controlling step for the WTP Project is vitrification of the large volume of LAW, current capacity of the WTP (as planned) would require 50 years to treat the Hanford tank waste, if the entire LAW volume were to be processed through the WTP. To reduce the time and cost for treatment of Hanford Tank Waste, and as required by the Tank Waste Remediation System Environmental Impact Statement Record of Decision and the Hanford Federal Facility Consent Agreement (Tn-Party Agreement), DOE plans to supplement the LAW treatment capacity of the WTP. Since 2002, DOE, in cooperation with the Environmental Protection Agency and State of Washington Department of Ecology has been evaluating technologies that could provide safe and effective supplemental treatment of LAW. Current efforts at Hanford are intended to provide additional information to aid a joint agency decision on which technology will be used to supplement the WTP. A Research, Development and Demonstration permit has been issued by the State of Washington to build and (for a limited time) operate a Demonstration Bulk Vitrification System (DBVS) facility to provide information for the decision on a supplemental treatment technology for up to 50% of the LAW. In the Bulk Vitrification (BV) process, LAW, soil, and glass-forming chemicals are mixed, dried, and placed in a refractory-lined box, Electric current, supplied through two graphite electrodes in the box, melts the waste feed, producing a durable glass waste-form. Although recent modifications to the process have resulted in significant improvements, there are continuing technical concerns.

  5. Biogenic Hydroxyapatite: A New Material for the Preservation and Restoration of the Built Environment.

    Science.gov (United States)

    Turner, Ronald J; Renshaw, Joanna C; Hamilton, Andrea

    2017-09-20

    Ordinary Portland cement (OPC) is by weight the world's most produced man-made material and is used in a variety of applications in environments ranging from buildings, to nuclear wasteforms, and within the human body. In this paper, we present for the first time the direct deposition of biogenic hydroxyapatite onto the surface of OPC in a synergistic process which uses the composition of the cement substrate. This hydroxyapatite is very similar to that found in nature, having a similar crystallite size, iron and carbonate substitution, and a semi-crystalline structure. Hydroxyapatites with such a structure are known to be mechanically stronger and more biocompatible than synthetic or biomimetic hydroxyapatites. The formation of this biogenic hydroxyapatite coating therefore has significance in a range of contexts. In medicine, hydroxyapatite coatings are linked to improved biocompatibility of ceramic implant materials. In the built environment, hydroxyapatite coatings have been proposed for the consolidation and protection of sculptural materials such as marble and limestone, with biogenic hydroxyapatites having reduced solubility compared to synthetic apatites. Hydroxyapatites have also been established as effective for the adsorption and remediation of environmental contaminants such as radionuclides and heavy metals. We identify that in addition to providing a biofilm scaffold for nucleation, the metabolic activity of Pseudomonas fluorescens increases the pH of the growth medium to a suitable level for hydroxyapatite formation. The generated ammonia reacts with phosphate in the growth medium, producing ammonium phosphates which are a precursor to the formation of hydroxyapatite under conditions of ambient temperature and pressure. Subsequently, this biogenic deposition process takes place in a simple reaction system under mild chemical conditions and is cheap and easy to apply to fragile biological or architectural surfaces.

  6. ENGINEERED BARRIER SYSTEM FEATURES, EVENTS AND PROCESSES

    Energy Technology Data Exchange (ETDEWEB)

    Jaros, W.

    2005-08-30

    The purpose of this report is to evaluate and document the inclusion or exclusion of engineered barrier system (EBS) features, events, and processes (FEPs) with respect to models and analyses used to support the total system performance assessment for the license application (TSPA-LA). A screening decision, either Included or Excluded, is given for each FEP along with the technical basis for exclusion screening decisions. This information is required by the U.S. Nuclear Regulatory Commission (NRC) at 10 CFR 63.114 (d, e, and f) [DIRS 173273]. The FEPs addressed in this report deal with those features, events, and processes relevant to the EBS focusing mainly on those components and conditions exterior to the waste package and within the rock mass surrounding emplacement drifts. The components of the EBS are the drip shield, waste package, waste form, cladding, emplacement pallet, emplacement drift excavated opening (also referred to as drift opening in this report), and invert. FEPs specific to the waste package, cladding, and drip shield are addressed in separate FEP reports: for example, ''Screening of Features, Events, and Processes in Drip Shield and Waste Package Degradation'' (BSC 2005 [DIRS 174995]), ''Clad Degradation--FEPs Screening Arguments (BSC 2004 [DIRS 170019]), and Waste-Form Features, Events, and Processes'' (BSC 2004 [DIRS 170020]). For included FEPs, this report summarizes the implementation of the FEP in the TSPA-LA (i.e., how the FEP is included). For excluded FEPs, this analysis provides the technical basis for exclusion from TSPA-LA (i.e., why the FEP is excluded). This report also documents changes to the EBS FEPs list that have occurred since the previous versions of this report. These changes have resulted due to a reevaluation of the FEPs for TSPA-LA as identified in Section 1.2 of this report and described in more detail in Section 6.1.1. This revision addresses updates in Yucca Mountain Project

  7. Evaluation of Technetium Getters to Improve the Performance of Cast Stone

    Energy Technology Data Exchange (ETDEWEB)

    Neeway, James J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Qafoku, Nikolla P. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Serne, R. Jeffrey [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lawter, Amanda R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Stephenson, John R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lukens, Wayne W. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Westsik, Joseph H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-11-01

    Cast Stone has been selected as the preferred waste form for solidification of aqueous secondary liquid effluents from the Hanford Tank Waste Treatment and Immobilization Plant (WTP) process condensates and low-activity waste (LAW) melter off-gas caustic scrubber effluents. Cast Stone is also being evaluated as a supplemental immobilization technology to provide the necessary LAW treatment capacity to complete the Hanford tank waste cleanup mission in a timely and cost effective manner. One of the major radionuclides that Cast Stone has the potential to immobilize is technetium (Tc). The mechanism for immobilization is through the reduction of the highly mobile Tc(VII) species to the less mobile Tc(IV) species by the blast furnace slag (BFS) used in the Cast Stone formulation. Technetium immobilization through this method would be beneficial because Tc is one of the most difficult contaminants to address at the U.S. Department of Energy (DOE) Hanford Site due to its complex chemical behavior in tank waste, limited incorporation in mid- to high-temperature immobilization processes (vitrification, steam reformation, etc.), and high mobility in subsurface environments. In fact, the Tank Closure and Waste Management Environmental Impact Statement for the Hanford Site, Richland, Washington (TC&WM EIS) identifies technetium-99 (99Tc) as one of the radioactive tank waste components contributing the most to the environmental impact associated with the cleanup of the Hanford Site. The TC&WM EIS, along with an earlier supplemental waste-form risk assessment, used a diffusion-limited release model to estimate the release of different contaminants from the WTP process waste forms. In both of these predictive modeling exercises, where effective diffusivities based on grout performance data available at the time, groundwater at the 100-m down-gradient well exceeded the allowable maximum permissible concentrations for 99Tc. (900 pCi/L). Recent relatively

  8. Synthesis, structure elucidation and redox properties of 99Tc complexes of lacunary Wells Dawson polyoxometalates: insights into molecular 99Tc - metal oxide interactions

    Energy Technology Data Exchange (ETDEWEB)

    McGregor, Donna; Burton-Pye, Benjamin P.; Howell, Robertha C.; Mbomekalle, Israel M.; Lukens Jr, Wayne W.; Bian, Fang; Mausolf, Edward; Poineau, Frederic; Czerwinski, Kenneth R; Francesconi, Lynn C.

    2011-01-10

    The isotope 99Tc (beta max: 250 keV, half-life: 2 x 105 year) is an abundant product of uranium-235 fission in nuclear reactors and is present throughout the radioactive waste stored in underground tanks at Hanford and Savannah River. Understanding and controlling the extensive redox chemistry of 99Tc is important to identify tunable strategies to separate 99Tc from spent fuel and from waste tanks and once separated, to identify and develop an appropriately stable waste-form for 99Tc. Polyoxometalates (POMs), nanometer sized models for metal oxide solid-state materials, are used in this study to provide a molecular level understanding of the speciation and redox chemistry of incorporated 99Tc. In this study, 99Tc complexes of the (alpha 2-P2W17O61)10- and (alpha 1-P2W17O61)10- isomers were prepared. Ethylene glycol was used as a"transfer ligand" to minimize the formation of TcO2 cdot xH2O. The solution structures, formulations, and purity of TcVO(alpha 1/alpha 2-P2W17O61)7- were determined by multinuclear NMR. X-ray Absorption Spectroscopy of the complexes are in agreement with the formulation and structures determined from 31P and 183W NMR. Preliminary electrochemistry results are consistent with the EXAFS results, showing a facile reduction of the TcVO(alpha 1-P2W17O61)7- species compared to the TcVO(alpha 2-P2W17O61)7- analog. The alpha1- defect is unique in that a basic oxygen atom is positioned toward the alpha1- site and the TcVO center appears to form a dative metal-metal bond with a framework W site. These attributes may lead to the assistance of protonation events that facilitate reduction. Electrochemistry comparison shows that the ReV analogs are about 200 mV more difficult to reduce in accordance with periodic trends.

  9. The Crystal Structure and Chemistry of Pyrochlore-and Hollandite-Type Minerals and Their Application as Nuclear Waste-Forms%烧绿石及碱硬锰矿型矿物晶体化学及其核废料固化基材研究进展

    Institute of Scientific and Technical Information of China (English)

    李国武; 邢晓琳; 徐凯

    2016-01-01

    Pyrochlore- and hollandite-type minerals were extensively investigated as the nuclear waste-forms, showing the unique properties on thermal and chemical stabilities and irradiation resistance. In the past decade, our group methodically studied the crystal structure and chemistry of pyrochlore- and holladnite-type natural minerals and their synthesized compounds. According to X-ray diffraction analysis of fourteen synthesized rare-earth pyrochlore (Ree2B2O7, B=Ti, Zr), both Ree2Ti2O7 and Ree2Zr2O7 show obvious crystal structure variable. The study of hollandite crystal structure indicates that contents of cationA (A0-2B8O16,A=Na, K, Sr, Ba, etc.,B=Ti, Fe, Mn, Al, etc.) in the tunnel and parameters regarding anisotropy usually cause one-dimension incommensurate modulation structure along with c direction. This paper summarized our recent results on their structure and chemistry, and reviewed the research and development of their application in nuclear waste immobilization.%研究表明烧绿石和碱硬锰矿在固化核废料方面具有良好的性能。近年来我们对烧绿石型和碱硬锰矿型天然矿物以及系列人工合成矿物的晶体结构及晶体化学进行了系统详细的研究。通过人工合成14种镧系稀土的Ti和Zr的Ree2B2O7系列氧化物的X射线衍射实验,发现其Ti的Ree2Ti2O7系列氧化物和Zr的Ree2Zr2O7系列氧化物都出现了明显的晶变现象。晶体结构研究显示碱硬锰矿型结构矿物中 A 类阳离子在孔道中的含量变化及各向异性参数的变化导致了该型结构往往存在沿 c轴方向的一维非公度调制结构。同时对烧绿石和碱硬锰矿在固化核废料方面的研究现状及进展进行了综述。

  10. New Fission-Product Waste Forms: Development and Characterization

    Energy Technology Data Exchange (ETDEWEB)

    Alexandra Navrotsky

    2010-07-30

    Research performed on the program “New Fission Product Waste Forms: Development and Characterization,” in the last three years has fulfilled the objectives of the proposal which were to 1) establish ceramic waste forms for disposing of Cs, Sr and minor actinides, 2) fully characterize the phase relationships, structures and thermodynamic and kinetic stabilities of promising waste forms, 3) establish a sound technical basis for understanding key waste form properties, such as melting temperatures and aqueous durability, based on an in-depth understanding of waste form structures and thermochemistry, and 4) establish synthesis, testing, scaleup and commercialization routes for wasteform implementation through out in-kind collaborations. In addition, since Cs and Sr form new elements by radioactive decay, the behavior and thermodynamics of waste forms containing different proportions of Cs, Sr and their decay products were discovered using non-radioactive analogues. Collaborations among researchers from three institutions, UC Davis, Sandia National Laboratories, and Shott Inc., were formed to perform the primary work on the program. The unique expertise of each of the members in the areas of waste form development, structure/property relationships, hydrothermal and high temperature synthesis, crystal/glass production, and thermochemistry was critical to program success. In addition, collaborations with the Brigham Young Univeristy, Ben Gurion University, and Los Alamos National Laboratory, were established for standard entropies of ceramic waste forms, sol-gel synthesis, and high temperature synthesis. This work has had a significant impact in a number of areas. First, the studies of the thermodynamic stability of the mineral analogues provided an important technical foundation for assessment the viability of multicomponent oxide phases for Cs and Sr removal. Moreover, the thermodynamic data discovered in this program established information on the reaction

  11. NRC Consultation and Monitoring at the Savannah River Site: Focusing Reviews of Two Different Disposal Actions - 12181

    Energy Technology Data Exchange (ETDEWEB)

    Ridge, A. Christianne; Barr, Cynthia S.; Pinkston, Karen E.; Parks, Leah S.; Grossman, Christopher J.; Alexander, George W. [U.S. Nuclear Regulatory Commission (United States)

    2012-07-01

    Section 3116 of the Ronald W. Reagan National Defense Authorization Act for Fiscal Year 2005 (NDAA) requires the U.S. Department of Energy (DOE) to consult with the U.S. Nuclear Regulatory Commission (NRC) for certain non-high level waste determinations. The NDAA also requires NRC to monitor DOE's disposal actions related to those determinations. In Fiscal Year 2011, the NRC staff reviewed DOE performance assessments for tank closure at the F-Tank Farm (FTF) Facility and salt waste disposal at the Saltstone Disposal Facility (SDF) at the Savannah River Site (SRS) as part of consultation and monitoring, respectively. Differences in inventories, waste forms, and key barriers led to different areas of focus in the NRC reviews of these two activities at the SRS. Because of the key role of chemically reducing grouts in both applications, the evaluation of chemical barriers was significant to both reviews. However, radionuclide solubility in precipitated metal oxides is expected to play a significant role in FTF performance whereas release of several key radionuclides from the SDF is controlled by sorption or precipitation within the cementitious wasteform itself. Similarly, both reviews included an evaluation of physical barriers to flow, but differences in the physical configurations of the waste led to differences in the reviews. For example, NRC's review of the FTF focused on the modeled degradation of carbon steel tank liners while the staff's review of the SDF performance included a detailed evaluation of the physical degradation of the saltstone wasteform and infiltration-limiting closure cap. Because of the long time periods considered (i.e., tens of thousands of years), the NRC reviews of both facilities included detailed evaluation of the engineered chemical and physical barriers. The NRC staff reviews of residual waste disposal in the FTF and salt waste disposal in the SDF focused on physical barriers to flow and chemical barriers to

  12. Defining And Characterizing Sample Representativeness For DWPF Melter Feed Samples

    Energy Technology Data Exchange (ETDEWEB)

    Shine, E. P.; Poirier, M. R.

    2013-10-29

    statisticians used carefully thought out designs that systematically and economically provided plans for data collection from the DWPF process. Key shared features of the sampling designs used at DWPF and the Gy sampling methodology were the specification of a standard for sample representativeness, an investigation that produced data from the process to study the sampling function, and a decision framework used to assess whether the specification was met based on the data. Without going into detail with regard to the seven errors identified by Pierre Gy, as excellent summaries are readily available such as Pitard [1989] and Smith [2001], SRS engineers understood, for example, that samplers can be biased (Gy's extraction error), and developed plans to mitigate those biases. Experiments that compared installed samplers with more representative samples obtained directly from the tank may not have resulted in systematically partitioning sampling errors into the now well-known error categories of Gy, but did provide overall information on the suitability of sampling systems. Most of the designs in this report are related to the DWPF vessels, not the large SRS Tank Farm tanks. Samples from the DWPF Slurry Mix Evaporator (SME), which contains the feed to the DWPF melter, are characterized using standardized analytical methods with known uncertainty. The analytical error is combined with the established error from sampling and processing in DWPF to determine the melter feed composition. This composition is used with the known uncertainty of the models in the Product Composition Control System (PCCS) to ensure that the wasteform that is produced is comfortably within the acceptable processing and product performance region. Having the advantage of many years of processing that meets the waste glass product acceptance criteria, the DWPF process has provided a considerable amount of data about itself in addition to the data from many special studies. Demonstrating representative

  13. Secondary Waste Form Down-Selection Data Package—DuraLith

    Energy Technology Data Exchange (ETDEWEB)

    Mattigod, Shas V.; Westsik, Joseph H.

    2011-09-15

    This data package developed for the DuraLith wasteform includes information available in the open literature and from data obtained from testing currently underway. DuraLith is an alkali-activated geopolymer waste form developed by the Vitreous State Laboratory at The Catholic University of America (VSL-CUA) for encapsulating liquid radioactive waste. A DuraLith waste form developed for treating Hanford secondary waste liquids is prepared by alkali-activation of a mixture of ground blast furnace slag and metakaolinite with sand used as a filler material. Based on optimization tests, solid waste loading of {approx}7.5% and {approx}14.7 % has been achieved using the Hanford secondary waste S1 and S4 simulants, respectively. The Na loading in both cases is equivalent to {approx}6 M. Some of the critical parameters for the DuraLith process include, hydrogen generation and heat evolution during activator solution preparation using the waste simulant, heat evolution during and after mixing the activator solution with the dry ingredients, and a working window of {approx}20 minutes to complete the pouring of the DuraLith mixture into molds. Results of the most recent testing indicated that the working window can be extended to {approx}30 minutes if 75 wt% of the binder components, namely, blast furnace slag and metakaolin are replaced by Class F fly ash. A preliminary DuraLith process flow sheet developed by VSL-CUA for processing Hanford secondary waste indicated that 10 to 22 waste monoliths (each 48 ft3 in volume) can be produced per day. There are no current pilot-scale or full-scale DuraLith plants under construction or in operation; therefore, the cost of DuraLith production is unknown. The results of the non-regulatory leach tests, EPA Draft 1313 and 1316, Waste Simulant S1-optimized DuraLith specimens indicated that the concentrations of RCRA metals (Ag, Cd, Cr, Hg, and Pb) in the leachates were well below the Universal Treatment Standard limits in 40 CFR 268

  14. Development of Advanced Electrochemical Emission Spectroscopy for Monitoring Corrosion in Simulated DOE Liquid Waste

    Energy Technology Data Exchange (ETDEWEB)

    Digby D. Macdonald; Brian M. Marx; Sejin Ahn; Julio de Ruiz; Balaji Soundararaja; Morgan Smith; and Wendy Coulson

    2008-01-15

    Various forms of general and localized corrosion represent principal threats to the integrity of DOE liquid waste storage tanks. These tanks, which are of a single wall or double wall design, depending upon their age, are fabricated from welded carbon steel and contain a complex waste-form comprised of NaOH and NaNO{sub 3}, along with trace amounts of phosphate, sulfate, carbonate, and chloride. Because waste leakage can have a profound environmental impact, considerable interest exists in predicting the accumulation of corrosion damage, so as to more effectively schedule maintenance and repair. The different tasks that are being carried out under the current program are as follows: (1) Theoretical and experimental assessment of general corrosion of iron/steel in borate buffer solutions by using electrochemical impedance spectroscopy (EIS), ellipsometry and XPS techniques; (2) Development of a damage function analysis (DFA) which would help in predicting the accumulation of damage due to pitting corrosion in an environment prototypical of DOE liquid waste systems; (3) Experimental measurement of crack growth rate, acoustic emission signals and coupling currents for fracture in carbon and low alloy steels as functions of mechanical (stress intensity), chemical (conductivity), electrochemical (corrosion potential, ECP), and microstructural (grain size, precipitate size, etc) variables in a systematic manner, with particular attention being focused on the structure of the noise in the current and its correlation with the acoustic emissions; (4) Development of fracture mechanisms for carbon and low alloy steels that are consistent with the crack growth rate, coupling current data and acoustic emissions; (5) Inserting advanced crack growth rate models for SCC into existing deterministic codes for predicting the evolution of corrosion damage in DOE liquid waste storage tanks; (6) Computer simulation of the anodic and cathodic activity on the surface of the steel samples

  15. RESULTS OF THE FY09 ENHANCED DOE HIGH LEVEL WASTE MELTER THROUGHPUT STUDIES AT SRNL

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, F.; Edwards, T.

    2010-06-23

    High-level waste (HLW) throughput (i.e., the amount of waste processed per unit time) is a function of two critical parameters: waste loading (WL) and melt rate. For the Waste Treatment and Immobilization Plant (WTP) at the Hanford Site and the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS), increasing HLW throughput would significantly reduce the overall mission life cycle costs for the Department of Energy (DOE). The objective of this task is to develop data, assess property models, and refine or develop the necessary models to support increased WL of HLW at SRS. It is a continuation of the studies initiated in FY07, but is under the specific guidance of a Task Change Request (TCR)/Work Authorization received from DOE headquarters (Project Number RV071301). Using the data generated in FY07, FY08 and historical data, two test matrices (60 glasses total) were developed at the Savannah River National Laboratory (SRNL) in order to generate data in broader compositional regions. These glasses were fabricated and characterized using chemical composition analysis, X-ray Diffraction (XRD), viscosity, liquidus temperature (TL) measurement and durability as defined by the Product Consistency Test (PCT). The results of this study are summarized below: (1) In general, the current durability model predicts the durabilities of higher waste loading glasses quite well. A few of the glasses exhibited poorer durability than predicted. (2) Some of the glasses exhibited anomalous behavior with respect to durability (normalized leachate for boron (NL [B])). The quenched samples of FY09EM21-02, -07 and -21 contained no nepheline or other wasteform affecting crystals, but have unacceptable NL [B] values (> 10 g/L). The ccc sample of FY09EM21-07 has a NL [B] value that is more than one half the value of the quenched sample. These glasses also have lower concentrations of Al{sub 2}O{sub 3} and SiO{sub 2}. (3) Five of the ccc samples (EM-13, -14, -15, -29 and

  16. 99Tc and Re incorporated into metal oxide polyoxometalates: oxidation state stability elucidated by electrochemistry and theory.

    Science.gov (United States)

    McGregor, Donna; Burton-Pye, Benjamin P; Mbomekalle, Israel M; Aparicio, Pablo A; Romo, Susanna; López, Xavier; Poblet, Josep M; Francesconi, Lynn C

    2012-08-20

    The radioactive element technetium-99 ((99)Tc, half-life = 2.1 × 10(5) years, β(-) of 253 keV), is a major byproduct of (235)U fission in the nuclear fuel cycle. (99)Tc is also found in radioactive waste tanks and in the environment at National Lab sites and fuel reprocessing centers. Separation and storage of the long-lived (99)Tc in an appropriate and stable waste-form is an important issue that needs to be addressed. Considering metal oxide solid-state materials as potential storage matrixes for Tc, we are examining the redox speciation of Tc on the molecular level using polyoxometalates (POMs) as models. In this study we investigate the electrochemistry of Tc complexes of the monovacant Wells-Dawson isomers, α(1)-P(2)W(17)O(61)(10-) (α1) and α(2)-P(2)W(17)O(61)(10-) (α2) to identify features of metal oxide materials that can stabilize the immobile Tc(IV) oxidation state accessed from the synthesized Tc(V)O species and to interrogate other possible oxidation states available to Tc within these materials. The experimental results are consistent with density functional theory (DFT) calculations. Electrochemistry of K(7-n)H(n)[Tc(V)O(α(1)-P(2)W(17)O(61))] (Tc(V)O-α1), K(7-n)H(n)[Tc(V)O(α(2)-P(2)W(17)O(61))] (Tc(V)O-α2) and their rhenium analogues as a function of pH show that the Tc-containing derivatives are always more readily reduced than their Re analogues. Both Tc and Re are reduced more readily in the lacunary α1 site as compared to the α2 site. The DFT calculations elucidate that the highest oxidation state attainable for Re is VII while, under the same electrochemistry conditions, the highest oxidation state for Tc is VI. The M(V)→ M(IV) reduction processes for Tc(V)O-α1 are not pH dependent or only slightly pH dependent suggesting that protonation does not accompany reduction of this species unlike the M(V)O-α2 (M = (99)Tc, Re) and Re(V)O-α1 where M(V/IV) reduction process must occur hand in hand with protonation of the terminal M═O to

  17. SLURRY MIX EVAPORATOR BATCH ACCEPTABILITY AND TEST CASES OF THE PRODUCT COMPOSITION CONTROL SYSTEM WITH THORIUM AS A REPORTABLE ELEMENT

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, T.

    2010-10-07

    The Defense Waste Processing Facility (DWPF), which is operated by Savannah River Remediation, LLC (SRR), has recently begun processing Sludge Batch 6 (SB6) by combining it with Frit 418 at a nominal waste loading (WL) of 36%. A unique feature of the SB6/Frit 418 glass system, as compared to the previous glass systems processed in DWPF, is that thorium will be a reportable element (i.e., concentrations of elemental thorium in the final glass product greater than 0.5 weight percent (wt%)) for the resulting wasteform. Several activities were initiated based upon this unique aspect of SB6. One of these was an investigation into the impact of thorium on the models utilized in DWPF's Product Composition and Control System (PCCS). While the PCCS is described in more detail below, for now note that it is utilized by Waste Solidification Engineering (WSE) to evaluate the acceptability of each batch of material in the Slurry Mix Evaporator (SME) before this material is passed on to the melter. The evaluation employs models that predict properties associated with processability and product quality from the composition of vitrified samples of the SME material. The investigation of the impact of thorium on these models was conducted by Peeler and Edwards [1] and led to a recommendation that DWPF can process the SB6/Frit 418 glass system with ThO{sub 2} concentrations up to 1.8 wt% in glass. Questions also arose regarding the handling of thorium in the SME batch acceptability process as documented by Brown, Postles, and Edwards [2]. Specifically, that document is the technical bases of PCCS, and while Peeler and Edwards confirmed the reliability of the models, there is a need to confirm that the current implementation of DWPF's PCCS appropriately handles thorium as a reportable element. Realization of this need led to a Technical Task Request (TTR) prepared by Bricker [3] that identified some specific SME-related activities that the Savannah River National Laboratory

  18. Defining And Characterizing Sample Representativeness For DWPF Melter Feed Samples

    Energy Technology Data Exchange (ETDEWEB)

    Shine, E. P.; Poirier, M. R.

    2013-10-29

    statisticians used carefully thought out designs that systematically and economically provided plans for data collection from the DWPF process. Key shared features of the sampling designs used at DWPF and the Gy sampling methodology were the specification of a standard for sample representativeness, an investigation that produced data from the process to study the sampling function, and a decision framework used to assess whether the specification was met based on the data. Without going into detail with regard to the seven errors identified by Pierre Gy, as excellent summaries are readily available such as Pitard [1989] and Smith [2001], SRS engineers understood, for example, that samplers can be biased (Gy's extraction error), and developed plans to mitigate those biases. Experiments that compared installed samplers with more representative samples obtained directly from the tank may not have resulted in systematically partitioning sampling errors into the now well-known error categories of Gy, but did provide overall information on the suitability of sampling systems. Most of the designs in this report are related to the DWPF vessels, not the large SRS Tank Farm tanks. Samples from the DWPF Slurry Mix Evaporator (SME), which contains the feed to the DWPF melter, are characterized using standardized analytical methods with known uncertainty. The analytical error is combined with the established error from sampling and processing in DWPF to determine the melter feed composition. This composition is used with the known uncertainty of the models in the Product Composition Control System (PCCS) to ensure that the wasteform that is produced is comfortably within the acceptable processing and product performance region. Having the advantage of many years of processing that meets the waste glass product acceptance criteria, the DWPF process has provided a considerable amount of data about itself in addition to the data from many special studies. Demonstrating representative

  19. Coupled thermo-hydro-chemical models of swelling bentonites

    Science.gov (United States)

    Samper, Javier; Mon, Alba; Zheng, Liange; Montenegro, Luis; Naves, Acacia; Pisani, Bruno

    2014-05-01

    The disposal of radioactive waste in deep geological repositories is based on the multibarrier concept of retention of the waste by a combination of engineered and geological barriers. The engineered barrier system (EBS) includes the solid conditioned waste-form, the waste container, the buffer made of materials such as clay, grout or crushed rock that separate the waste package from the host rock and the tunnel linings and supports. The geological barrier supports the engineered system and provides stability over the long term during which time radioactive decay reduces the levels of radioactivity. The strong interplays among thermal (T), hydrodynamic (H), mechanical (M) and chemical (C) processes during the hydration, thermal and solute transport stages of the engineered barrier system (EBS) of a radioactive waste repository call for coupled THMC models for the metallic overpack, the unsaturated compacted bentonite and the concrete liner. Conceptual and numerical coupled THMC models of the EBS have been developed, which have been implemented in INVERSE-FADES-CORE. Chemical reactions are coupled to the hydrodynamic processes through chemical osmosis (C-H coupling) while bentonite swelling affects solute transport via changes in bentonite porosity changes (M-H coupling). Here we present THMC models of heating and hydration laboratory experiments performed by CIEMAT (Madrid, Spain) on compacted FEBEX bentonite and numerical models for the long-term evolution of the EBS for 1 Ma. The changes in porosity caused by swelling are more important than those produced by the chemical reactions during the early evolution of the EBS (t < 100 years). For longer times, however, the changes in porosity induced by the dissolution/precipitation reactions are more relevant due to: 1) The effect of iron mineral phases (corrosion products) released by the corrosion of the carbon steel canister; and 2) The hyper alkaline plume produced by the concrete liner. Numerical results show that

  20. Ion Exchange Distribution Coefficient Tests and Computer Modeling at High Ionic Strength Supporting Technetium Removal Resin Maturation

    Energy Technology Data Exchange (ETDEWEB)

    Nash, Charles A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hamm, L. Larry [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Smith, Frank G. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); McCabe, Daniel J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2014-12-19

    stream without dilution and to minimize the volume of the final wasteform. This work examined the impact of high ionic strength, high density, and high viscosity if higher concentration LAW feed solution is used. Perrhenate (ReO4-) has been shown to be a good nonradioactive surrogate for pertechnetate in laboratory testing for this ion exchange resin, and the performance bias is well established. Equilibrium contact testing with 7.8 M [Na+] average simulant concentrations indicated that the SuperLig® 639 resin average perrhenate distribution coefficient was 368 mL/g at a 100:1 phase ratio. Although this indicates good performance at high ionic strength, an equilibrium test cannot examine the impact of liquid viscosity, which impacts the diffusivity of ions and therefore the loading kinetics. To get an understanding of the effect of diffusivity, modeling was performed, which will be followed up with column tests in the future.

  1. Sintered bentonite ceramics for the immobilization of cesium- and strontium-bearing radioactive waste

    Science.gov (United States)

    Ortega, Luis Humberto

    were also tested. The final solid product was a hard dense ceramic with a density that varied from 2.12 g/cm3 for a 19% waste loading with a 1200°C sintering temperature to 3.03 g/cm 3 with a 29% waste loading and sintered at 1100°C. Differential Scanning Calorimetry and Thermal Gravimetric Analysis (DSC-TGA) of the loaded bentonite displayed mass loss steps which were consistent with water losses in pure bentonite. Water losses were complete after dehydroxylation at ˜650°C. No mass losses were evident beyond the dehydroxylation. The ceramic melts at temperatures greater than 1300°C. Light flash analysis found heat capacities of the ceramic to be comparable to those of strontium and barium feldspars as well as pollucite. Thermal conductivity improved with higher sintering temperatures, attributed to lower porosity. Porosity was minimized in 1200°C sinterings. Ceramics with waste loadings less than 25 wt% displayed slump, the lowest waste loading, 15 wt% bloated at a 1200°C sintering. Waste loading above 25 wt% produced smooth uniform ceramics when sintered >1100°C. Sintered bentonite may provide a simple alternative to vitrification and other engineered radioactive waste-forms.

  2. Scientific Analysis Cover Sheet for Radionuclide Screening

    Energy Technology Data Exchange (ETDEWEB)

    G. Ragan

    2002-08-09

    pressurized water reactor (PWR) fuel, U.S. Department of Energy (DOE) spent nuclear fuel (DSNF), and high-level waste (HLW). Average and outlying (high burnup, high initial enrichment, low age, or otherwise exceptional) forms of each waste-form type are considered. This analysis has been prepared in accordance with a technical work plan (BSC 2002c). In a review of Revision 00 of this radionuclide screening analysis, the NRC found that ''processes that affect transport in the biosphere, such as uptake by plants and bioaccumulation are not accounted for'' and that ''the direct exposure pathway is not accounted for'' (Beckman 2001, Section 5.3.2.1). The NRC also found that the solubility and sorption classes were too broadly defined, noting, for example, that Se is in the same solubility and sorptivity groups as Np and U, yet is ''more soluble than Np and U by several orders of magnitude'' (Beckman 2001, Section 5.3.2.1). This revision seeks to build upon the strengths of the earlier screening method while responding to the specific concerns raised by the NRC and other reviewers. In place of simple inhalation and ingestion dose conversion factors, the revised radionuclide screening uses screening factors that also take into account soil accumulation, uptake by plants, exposure to contaminated ground, and other features of the biosphere that were neglected in the previous screening. Whereas the previous screening analysis allowed only two solubility classes (soluble and insoluble), the revised screening introduces an intermediate solubility class to better segregate the radionuclides into transport groups.

  3. Hanford Site Secondary Waste Roadmap

    Energy Technology Data Exchange (ETDEWEB)

    Westsik, Joseph H.

    2009-01-29

    performance requirements, waste composition, preliminary waste form screening, waste form development, process design and support, and validation. The regulatory and performance requirements activity will provide the secondary waste-form performance requirements. The waste-composition activity will provide workable ranges of secondary waste compositions and formulations for simulants and surrogates. Preliminary waste form screening will identify candidate waste forms for immobilizing the secondary wastes. The waste form development activity will mature the waste forms, leading to a selected waste form(s) with a defensible understanding of the long-term release rate and input into the critical decision process for a secondary waste treatment process/facility. The process and design support activity will provide a reliable process flowsheet and input to support a robust facility design. The validation effort will confirm that the selected waste form meets regulatory requirements. The final outcome of the implementation of the secondary waste roadmap is the compliant, effective, timely, and cost-effective disposal of the secondary wastes. The work necessary to address the programmatic, regulatory, and technical risks and uncertainties identified through the Secondary Waste Roadmap Workshop are assembled into several program needs elements. Programmatic/Regulatory needs include: • Select and deploy Hanford tank waste supplemental treatment technology • Provide treatment capability for secondary waste streams from tank waste treatment • Develop consensus on secondary waste form acceptance. Technology needs include: • Define secondary waste composition ranges and uncertainties • Identify and develop waste forms for secondary waste immobilization and disposal • Develop test methods to characterize secondary waste form performance. Details for each of these program elements are provided.

  4. Development of a technical approach for assessing environmental release and migration characteristics of Hanford Grout

    Energy Technology Data Exchange (ETDEWEB)

    Serne, R. J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Treat, R. L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Lokken, R. O. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    1985-09-01

    Hanford grout will not change significantly in its chemical nature once the major chemical reactions at the waste-form sediment interface are completed. Also, the range of sediments at Hanford through which the leachate will travel probably will not exhibit widely varying adsorption properties. These sediments are generally alkaline sands and silts containing little organic matter and have low-to-medium cation exchange capacities. Their interaction with the expected leachates from the Hanford grout should not appreciably affect the composition of the major constituents of the leachates. Therefore, the constant Rd adsorption model should be a useful first approximation of the adsorption processes likely to control trace concentrations of waste radionuclides and hazardous inorganic chemicals that may leach into the groundwater. Because the Rd approach is empirical, it does not lend itself to the identification of transport-controlling mechanisms, a key need for gaining credibility in longterm performance assessments. Despite its limitations, the Rd concept is believed to be a practical and useful tool for quantifying the interaction of Hanford grout leachate with Hanford sediments and assessing the mobility of waste species. Unlike waste-form leaching, the research of radionuclide adsorption does not have a programmatic focal point in which standardization of techniques and procedures is occurring. At present we recommend that sever a 1 different types of adsorption experiments be performed, including hatch and column tests. Both types of tests are needed to increase the probability that the deficiencies of each are addressed. The separation of the complex chemical interactions of grout, sediment and groundwater into simple leaching and adsorption processes for ease of experimentation and modeling is under question. Few experimenters have performed combined tests involving the waste form, sediment and leaching solution though such a combination represents the actual

  5. Recent outputs of the Oklo (Gabon) natural analogue study to nuclear waste disposal; Apports recents de l'etude de l'analogue naturel Oklo (Gabon) dans le domaine du stockage des dechets nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    Michaud, V.; Trotignon, L. [CEA Cadarache, Dept. d' Entreposage et de Stockage des Dechets (DCC/DESD/SESD), 13 - Saint-Paul-lez-Durance (France); Louvat, D. [CEA Cadarache, (IPSN/DPRE/SERNAT/LERCM), 13 - Saint-Paul-lez-Durance (France)

    2000-07-01

    nuclear waste disposal, the study of the long-term evolution of spent fuel and the long-term behavior of geological materials with respect to the containment of actinides and fission products. The Oklo natural analogue displays a number of specific features that make it unique in the world. The Oklo basin is characterized by the occurrence of meter scale uraninite lenses, that were affected by nuclear fission 2 billion years ago. These ''reactor zones'' exist in three sites: Oklo, Okelobondo and Bagombe. By analogy with a repository system, they are considered as representative of the 'Source' term. Numerous isotopic and geochemical tracers are thus available in order to restrict the migration or retention processes of actinides and fission products present in these zones. The near environment of the reactor zones, called ''Near field'' by analogy, is mainly composed of clayey materials (i.e. chlorite, illite, kaolinite). Reactor zones are found at present from the surface (Bagombe under oxidizing and acid conditions, with supergene weathering) to deep (Okelobondo under reducing conditions, with a low groundwater dynamics) conditions. Some reactor zones, e.g. R.Z. 13 in Oklo mine, have been subjected to strong hydrothermal disturbances (with temperatures above 350 deg C), linked to the geological history of the Franceville basin. On the other hand, the old age of the Oklo reactors (2 Ga) implies that pressure, temperature and chemical conditions have evolved during a long geological history, with associated basin scale movements of fluids. The Oklo-natural analogue Phase II project compiled useful information and tools for persons involved in Performance Assessment of waste disposal, wasteform conception or long term behavior [10] in four main areas corresponding to major investigation fields: 1/ ''Source'' term evolution, 2/ Long term containment properties of geological materials, 3/ Migration and