Search NASA⌕ Search

SEARCH · Search NASA

Results for “TECHNETIUM 99”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 235 records · Page 13

Waste Treatment and Immobilization Plant U. S. Department of Energy Office of River Protection Submerged Bed Scrubber Condensate Disposition Project - 13460

The Hanford Waste Treatment and Immobilization Plant (WTP) will generate an off-gas treatment system secondary liquid waste stream [submerged bed scrubber (SBS) condensate], which is currently planned for recycle back to the WTP Low Activity Waste (LAW) melter. This SBS condensate waste stream is high in Tc-99, which is not efficiently captured in the vitrified glass matrix [1]. A pre-conceptual engineering study was prepared in fiscal year 2012 to evaluate alternate flow paths for melter off-gas secondary liquid waste generated by the WTP LAW facility [2]. This study evaluated alternatives for direct off-site disposal of this SBS without pre-treatment, which mitigates potential issues associated with recycling. This study [2] concluded that SBS direct disposal is a viable option to the WTP baseline. The results show: - Off-site transportation and disposal of the SBS condensate is achievable and cost effective. - Reduction of approximately 4,325 vitrified WTP Low Activity Waste canisters could be realized. - Positive WTP operational impacts; minimal WTP construction impacts are realized. - Reduction of mass flow from the LAW Facility to the Pretreatment Facility by 66%. - Improved Double Shell Tank (DST) space management is a benefit. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Radiation Stability of Benzyl Tributyl Ammonium Chloride towards Technetium-99 Extraction - 13016

A closed nuclear fuel cycle combining new separation technologies along with generation III and generation IV reactors is a promising way to achieve a sustainable energy supply. But it is important to keep in mind that future recycling processes of used nuclear fuel (UNF) must minimize wastes, improve partitioning processes, and integrate waste considerations into processes. New separation processes are being developed worldwide to complement the actual industrialized PUREX process which selectively separates U(VI) and Pu(IV) from the raffinate. As an example, the UREX process has been developed in the United States to co-extract hexavalent uranium (U) and hepta-valent technetium (Tc) by tri-n-butyl phosphate (TBP). Tc-99 is recognized to be one of the most abundant, long-lived radio-toxic isotopes in UNF (half-life, t{sub 1/2} = 2.13 x 10{sup 5} years), and as such, is targeted in UNF separation strategies for isolation and encapsulation in solid waste-forms for final disposal in a nuclear waste repository. Immobilization of Tc-99 by a durable solid waste-form is a challenge, and its fate in new advanced technology processes is of importance. It is essential to be able to quantify and locate 1) its occurrence in any new developed flowsheets, 2) its chemical form in the individual phases of a process, 3) its potential quantitative transfer in any waste streams, and consequently, 4) its quantitative separation for either potential transmutation to Ru-100 or isolation and encapsulation in solid waste-forms for ultimate disposal. In addition, as a result of an U(VI)-Tc(VII) co-extraction in a UREX-based process, Tc(VII) could be found in low level waste (LLW) streams. There is a need for the development of new extraction systems that would selectively extract Tc-99 from LLW streams and concentrate it for feed into high level waste (HLW) for either Tc-99 immobilization in metallic waste-forms (Tc-Zr alloys), and/or borosilicate-based waste glass. Studies have been launched to investigate the suitability of new macro-compounds such as crown-ethers, aza-crown ethers, quaternary ammonium salts, and resorcin-arenes for the selective extraction of Tc-99 from nitric acid solutions. The selectivity of the ligand is important in evaluating potential separation processes and also the radiation stability of the molecule is essential for minimization of waste and radiolysis products. In this paper, we are reporting the extraction of TcO{sub 4}{sup -} by benzyl tributyl ammonium chloride (BTBA). Experimental efforts were focused on determining the best extraction conditions by varying the ligand's matrix conditions and concentration, as well as varying the organic phase composition (i.e. diluent variation). Furthermore, the ligand has been investigated for radiation stability. The ?-irradiation was performed on the neat organic phases containing the ligand at different absorbed doses to a maximum of 200 kGy using an external Co-60 source. Post-irradiation solvent extraction measurements will be discussed. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Startup of the New 200 West Pump-and-Treat, Hanford Site, Richland, Washington - 13214

On June 28, 2012, CH2M HILL Plateau Remediation Company (CHPRC) completed the construction and acceptance testing for a new 2,500 gallon-per-minute (gpm) pump-and-treat (P and T) system in the 200 West Area of the Hanford Site in Washington State. This system is designed to remove Tc-99, carbon tetrachloride, trichloroethene (TCE), nitrate, and total and hexavalent chromium from groundwater using ion exchange, anoxic and aerobic bioreactors, and air stripping. The system will eventually remove uranium from groundwater using ion exchange as well. The startup of the P and T system is important because it will ensure that contaminants from the 200 West Area never reach the Columbia River. When fully operational, the 200 West P and T will include approximately 23 extraction wells and 21 injection wells. The extraction wells are 8 inches in diameter, are completed with well screens 100 feet or more in length, and are distributed throughout the central portion of the 5-square-mile carbon tetrachloride plume. The injection wells are also 8 inches in diameter and are installed up-gradient of the plumes to recharge the aquifer and down-gradient of the plumes for flow-path control. Groundwater in the 200 West Area is approximately 250 feet below ground surface, and the aquifer is 200 feet or more in thickness. All of the contaminants (except nitrate) are found within the perimeter of the carbon tetrachloride plume and occur at various depths throughout the aquifer. The 200 West P and T consists of two separate buildings to conduct groundwater treatment. The RAD building contains an ion exchange system to remove Tc-99 from groundwater at a maximum flow rate of 600 gpm. The RAD building only accepts water from those extraction wells showing elevated Tc-99 concentrations. Groundwater initially fills an influent tank, is then pumped through particulate filters (to remove suspended materials), and then passes through two parallel treatment trains containing Purolite{sup R} A530E resin (which has been proven effective in removing Tc-99). The water is then transferred to the biological treatment building for further treatment. When the lead vessel in each of the two treatment trains becomes fully loaded with Tc-99, the Purolite A530E resin is transferred to a separate tank where it is heated to 160 deg. F to remove volatile organics prior to disposal at the Environmental Restoration Disposal Facility. The biological treatment building has a maximum flow capacity of 2,500 gpm. Groundwater from the nonradiological extraction wells and treated groundwater from the RAD building are initially pumped into an equalization tank and then into two parallel fluidized bed reactors (FBRs). The FBRs contain granulated activated carbon in suspension for microbes to populate, a carbon-based food source for the microbes to eat (e.g., MicroCg{sup TM}, molasses, or sodium lactate), and nitrate for the microbes to breathe (represents 'anoxic' conditions that contain little or no dissolved oxygen). The FBRs are maintained at a temperature between 55 deg. F and 90 deg. F, and at a pH between 6.5 and 6.8, to maximize microbial growth. The FBRs break down the nitrate, reduce the hexavalent chromium to trivalent chromium, and break down a good portion of the carbon tetrachloride and TCE. From the FBRs, groundwater is pumped through a carbon separation tank, then through a splitter box that divides the water evenly between four membrane bioreactors (MBRs) that further break down the contaminants. The MBRs have aeration capacity to provide sufficient oxygen for maintaining the aerobic biological process. The MBRs use submerged membranes for filtration. Vertically strung fibers are found in the membrane zone where a vacuum draws water through tiny pores in the fibers. The liquid is then pumped to air strippers to remove any volatile organics that have passed through the bioreactors. Solids from the MBRs are pumped to rotary drum thickeners and centrifuges for dewatering prior to lime being added to kill the bacteria and control odor. The conditioned sludge is then disposed at the Environmental Restoration Disposal Facility. The final treated water is then pumped to the injection well field. The 200 West P and T is a Leadership in Energy and Environmental Design (LEED) gold-certified facility that has an expected 70% energy cost savings over the lifetime of the plant. Fifty percent of the steel used in construction was recycled, 420 tons of recycled concrete were used, and over 75% of construction waste was diverted from landfills. The plant is only staffed one shift per day, and it uses a call-out system if the plant goes down during off-shift hours. Initial sampling results have shown that the 200 West P and T is successfully reducing contaminant levels well below the cleanup levels specified in the Record of Decision, Hanford 200 Area 200-ZP-1 Superfund Site, Benton County, Washington [1]. CHPRC and the U.S. Department of Energy are continuing to look for ways to reduce operational costs, such as testing alternative resins for removing radionuclides and finding less expensive carbon substrate sources. The benefit to others is a LEED gold-certified, 2,500 gpm P and T system that may be used as a model for other complex sites. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Fabrication and Properties of Technetium-Bearing Pyrochlores and Perovskites as Potential Waste Forms - 13222

Technetium-99 (t{sub 1/2}= 2.13x10{sup 5} years) is important from a nuclear waste perspective and is one of the most abundant, long-lived radioisotopes in used nuclear fuel (UNF). As such, it is targeted in UNF separation strategies such as UREX+, for isolation and encapsulation in solid waste forms for storage in a nuclear repository. We report here results regarding the incorporation of Tc-99 into ternary oxides of different structure types: pyrochlore (Nd{sub 2}Tc{sub 2}O{sub 7}), perovskite (SrTcO{sub 3}), and layered perovskite (Sr{sub 2}TcO{sub 4}). The goal was to determine synthesis conditions of these potential waste forms to immobilize Tc-99 as tetravalent technetium and to harvest crystallographic, thermophysical and hydrodynamic data. The objective of this research is to provide fundamental crystallographic and thermophysical data on advanced ceramic Tc-99 waste forms such as pyrochlore, perovskite, and layered perovskite in support of our current efforts on the corrosion of technetium-bearing waste forms. The ceramic Tc-99-bearing waste forms exhibit good crystallinity. The lattice parameters and crystal structures of the technetium host phases could be refined with high accuracies of ±3, ±4, and ±7 fm (10{sup -15} m), for Nd{sub 2}Tc{sub 2}O{sub 7}, SrTcO{sub 3}, and Sr{sub 2}TcO{sub 4}, respectively. The associated refinement residuals (R{sub Wp}) for the patterns are 4.1 %, 4.7 % and 6.7 %, and the refinement residuals for the individual phases (R{sub Bragg}) are 2.0 %, 2.4 % and 3.9 %, respectively. Thermophysical properties of the oxides SrTcO{sub 3}, Sr{sub 2}TcO{sub 4}, and Nd{sub 2}Tc{sub 2}O{sub 7} were analyzed using AC magnetic susceptibility measurements to further harvest information on the critical temperature (T{sub c}) for superconductivity. In our experiments the strontium technetates, SrTcO{sub 3} and Sr{sub 2}TcO{sub 4}, show superconductivity at rather high critical temperatures of T{sub c} = 7.8 K and 7 K, respectively. On the other hand Nd{sub 2}Tc{sub 2}O{sub 7} did not show any changes in magnetic properties above 3 K. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

The Cementitious Barriers Partnership (CBP) Software Toolbox Capabilities in Assessing the Degradation of Cementitious Barriers - 13487

The Cementitious Barriers Partnership (CBP) Project is a multi-disciplinary, multi-institutional collaboration supported by the U.S. Department of Energy (US DOE) Office of Tank Waste and Nuclear Materials Management. The CBP program has developed a set of integrated tools (based on state-of-the-art models and leaching test methods) that help improve understanding and predictions of the long-term structural, hydraulic and chemical performance of cementitious barriers used in nuclear applications. Tools selected for and developed under this program have been used to evaluate and predict the behavior of cementitious barriers used in near-surface engineered waste disposal systems for periods of performance up to 100 years and longer for operating facilities and longer than 1000 years for waste disposal. The CBP Software Toolbox has produced tangible benefits to the DOE Performance Assessment (PA) community. A review of prior DOE PAs has provided a list of potential opportunities for improving cementitious barrier performance predictions through the use of the CBP software tools. These opportunities include: 1) impact of atmospheric exposure to concrete and grout before closure, such as accelerated slag and Tc-99 oxidation, 2) prediction of changes in K{sub d}/mobility as a function of time that result from changing pH and redox conditions, 3) concrete degradation from rebar corrosion due to carbonation, 4) early age cracking from drying and/or thermal shrinkage and 5) degradation due to sulfate attack. The CBP has already had opportunity to provide near-term, tangible support to ongoing DOE-EM PAs such as the Savannah River Saltstone Disposal Facility (SDF) by providing a sulfate attack analysis that predicts the extent and damage that sulfate ingress will have on the concrete vaults over extended time (i.e., > 1000 years). This analysis is one of the many technical opportunities in cementitious barrier performance that can be addressed by the DOE-EM sponsored CBP software tools. Modification of the existing tools can provide many opportunities to bring defense in depth in prediction of the performance of cementitious barriers over time. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

The Transmutation of Nuclear Waste in the Two-Zone Subcritical System Driven by High- Intensity Neutron Generator - 12098

The main problems of transmutation of high-level radioactive waste (minor actinides and long-lived fission products) are considered in our work. The range of radioactive waste of nuclear power is analyzed. The conditions under which the transmutation of radioactive waste will be most effective are analyzed too. The modeling results of a transmutation of the main radioactive isotopes are presented and discussed. The transmutation of minor actinides and long-lived fission products are modeled in our work (minor actinides - Np-237, Am-241, Am-242, Am-243, Cm-244, Cm-245; long-lived fission products - I-129, Tc-99). The two-zone subcritical system is calculated with help of different neutron-physical codes (MCNP, Scale, Montebarn, Origen). The ENDF/B-VI nuclear data library used in above calculations. Thus, radioactive wastes can be divided into two main groups that need to be transmuted. The minor actinides form the first group and the long-lived fission products form the second one. For the purpose of effective transmutation these isotopes must be extracted from the spent nuclear fuel with the help of either PUREX technology or pyrometallurgical technology. The two-zone reactor system with fast and thermal regions is more effective for nuclear waste transmutation than the one-zone reactor. Modeling results show that nearly all radioactive wastes can be transmuted in the two-zone subcritical system driven by a high-intensity neutron generator with the external neutron source strength of 1.10{sup 13} n/sec. Obviously, transmutation rate will increase with a rise of the external neutron source strength. From the results above we can also see that the initial loading of radioactive isotopes into the reactor system should exceed by mass those isotopes that are finally produced. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Retention of Anionic Species on Granite: Influence of Granite Composition - 12129

Technetium (Tc-99, T{sub 1/2} = 2.1.10{sup 5} yrs) and selenium (Se-79, T{sub 1/2} = 6.5.10{sup 4} yrs) belong among fission products, being produced by fission of nuclear fuel. Both elements can significantly contribute to risk due to their complicated chemistry, long life times, high mobility and prevailing anionic character. Therefore, knowledge of migration behaviour under different conditions can significantly improve input into performance and safety assessment models. Granite is considered as a potential host rock for deep geological disposal of radioactive waste in many countries. Granitic rocks consist usually of quartz, feldspar, plagioclase (main components), mica, chlorite, kaolinite (minor components). The main feature of the rock is advection governed transport in fractures, complemented with diffusion process from fracture towards undisturbed rock matrix. The presented work is focused on interaction of anionic species (TcO{sub 4}{sup -}, SeO{sub 4}{sup 2-}, SeO{sub 3}{sup 2-}) with granitic rock. Furthermore, the importance of mineral composition on sorption of anionic species was also studied. The batch sorption experiments were conducted on the crushed granite from Bohemian Massive. Five fractions with defined grain size were used for static batch method. Mineral composition of each granitic fraction was evaluated using X-ray diffraction. The results showed differences in composition of granitic fractions, even though originating from one homogenized material. Sorption experiments showed influence of granite composition on adsorption of both TcO4{sup -} and SeO3{sup 2-} on granitic rock. Generally, Se(IV) showed higher retention than Tc(VII). Se(VI) was not almost sorbed at all. Fe containing minerals are pronounced as a selective Se and Tc sorbent, being reduced on their surface. As micas in granite are usually enriched in Fe, increased sorption of anionic species onto mica enriched fractions can be explained by this reason. On the other hand, fractions enriched in feldspar did not show increased sorption affinity to Tc and Se. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

K-25 D and D Challenges - 12170

Decontamination and decommissioning (D and D) of the K-25 former gaseous diffusion plant provides lessons learned applicable to other D and D projects in the DOE Complex. The 175,000- square-meter, 1.6-kilometer-long building is contaminated with highly enriched uranium (HEU), Tc-99, trace quantities of other fission products, asbestos, PCBs, and other hazardous wastes. Safety challenges include deteriorated electrical systems, significant structural degradation, and criticality and exposure risks. The project completely revised the D and D approach after a worker fell through an operating floor and was seriously injured. For protection from deteriorated building conditions, the project reduced the number of workers in the facility, limited their hours in the building, and installed nets and barriers to protect them from falls through weakened floors and from falling material. The new plan involved removing high-risk components, removing motors and compressors, and demolishing the building from the outside with heavy equipment with most of the piping and components inside the building during demolition. The team provided temporary electrical power; reconfigured the criticality alarm system; upgraded security; performed sampling and analysis to locate and characterize HEU deposits and Tc-99; and, to establish 'criticality incredible' conditions, conducted a nondestructive assay program and injected foam into equipment and piping as a contamination fixative. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Technical Basis for Gas-Phase Vadose Zone Remediation Technologies at Hanford: A Review - 12186

In situ vadose zone remediation approaches are being evaluated as potential options to mitigate the transport of inorganic and radionuclide contaminants from the vadose zone to the groundwater. Some of the candidate approaches are based on changing the contaminant or subsurface conditions in a way that slows downward migration of the contaminants through the vadose zone using amendments delivered in the gas-phase. Two promising approaches that have undergone testing at Hanford include soil desiccation to address technetium-99 contamination and ammonia-induced sequestration of uranium. For soil desiccation, a dry gas is injected to desiccate a targeted portion of the subsurface and thereby decrease contaminant movement by removing moisture and decreasing the hydraulic conductivity of the desiccated zone. Ammonia-induced sequestration of uranium relies on changing the pore water chemistry, primarily through pH changes, to induce dissolution and precipitation processes that decrease the amount of mobile uranium in the vadose zone. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Iron Phosphate Glass for Vitrifying Hanford AZ102 LAW in Joule Heated and Cold Crucible Induction Melters - 12240

An iron phosphate composition for vitrifying a high sulfate (∼17 wt%) and high alkali (∼80 wt%) Hanford low activity waste (LAW), known as AZ-102 LAW, has been developed for processing in a Joule Heated Melter (JHM) or a Cold Crucible Induction Melter (CCIM). This composition produced a glass waste form, designated as MS26AZ102F-2, with a waste loading of 26 wt% of the AZ-102 which corresponded to a total alkali and sulfate (represented as SO{sub 3}) content of 21 and 4.4 wt%, respectively. A slurry (7 M Na{sup +}) of MS26AZ102F-2 simulant was melted continuously at temperatures between 1030 and 1090 deg. C for 10 days in a small JHM at PNNL and for 70 hours in a CCIM at INL. The as-cast glasses produced in both melters and in trial laboratory experiments along with their canister centerline cooled (CCC) counterparts met the requirements for the Product Consistency Test (PCT) and the Vapor Hydration Test (VHT) responses in the Hanford Tank Waste Treatment and Immobilization Plant (WTP) Contract. These glass waste forms retained up to 77 % of the SO{sub 3} (3.3 wt%), 100% of the Cesium, and 33 to 44% of the rhenium (used as a surrogate for Tc) all of which either exceeded or were comparable to the retention limit for these species in borosilicate glass nuclear waste form. Analyses of commercial K-3 refractory lining and the Inconel 693 metal electrodes used in JHM indicated only minimum corrosion of these components by the iron phosphate glass. This is the first time that an iron phosphate composition was melted continuously in a slurry fed JHM and in the US, thereby, demonstrating that iron phosphate glasses can be used as alternative hosts for vitrifying nuclear waste. The following conclusions are drawn from the results of the present work. (1) An iron phosphate composition, designated as MS26AZ102F-2, containing 26 wt% of the simulated high sulfate (17 wt%), high alkali (80 wt%) Hanford AZ-102 LAW meets all the criteria for processing in a JHM and CCIM. This composition produces a homogeneous glass with a density of 2.80 ± 0.04 g/cm{sup 3} after melting between 1000 and 1050 deg. C for 3 to 5 h. (2) This is the first time that an iron phosphate glass was melted in the JHM continuously for 10 days achieving a specific melting rate of 1010 kg/m2/day and in the CCIM for 70 hrs with a melting rate of 664 kg/m{sup 2}/day. (3) The analyzed (ICP-AES) compositions for all the glasses prepared under several melting conditions are in excellent agreement with the target composition. The variation of melting conditions include: (i) use of small scale/short melting time to large scale/long melting time (300 g to 80 kg, 4 h to 10 days) operations including melting in the JHM and CCIM, (ii) use of dry or wet (slurry) melter feed, (iii) addition of reductant (sugar) in the batch, and (iv) bubbling the melt with air. (4) The chemical durability as measured by PCT and VHT for the quenched and CCC-treated waste forms prepared from laboratory, JHM or CCIM melting exceeds the DOE requirements for LAW. (5) Depending upon the melting time (4 h to 10 days), the average concentration of SO{sub 3} in the MS26AZ102F-2 iron phosphate glass waste form varies from about 1.78 (RSM) to 3.74 (laboratory melt) wt% which corresponds to an SO{sub 3} retention of 41 to 86% when melted between 1030 at 1050 deg. C in air. (6) The retention of SO{sub 3} in the glass was reduced when a reductant (sugar) was added to the slurry. (7) The retention of other problem components like Cesium and Rhenium/Tc-99 is also high; from 92 to 100% for Cs{sub 2}O under different melting conditions, and from 66 to 33% for Re{sub 2}O{sub 7} for melts processed from 3 to 5 hours and up to 10 days. (8) Corrosion tests on Inconel 693 and K-3 refractory at temperatures between 1000 and 1050 deg. C indicate that both materials should be suitable for melting iron phosphate glasses. (9) The viscosity and electrical conductivity of the MS26AZ102F-2 melt are within the acceptable limits for RSM and CCIM processing. (10)This work has shown that iron phosphate compositions are a potential alternative to borosilicate glasses for vitrifying problematic wastes such as Hanford AZ-102 LAW that contains high amounts sulfate and alkali oxides, and that iron phosphate melts such as MS26AZ102F-2 can be processed in the JHM or CCIM. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Radioactive Bench-scale Steam Reformer Demonstration of a Monolithic Steam Reformed Mineralized Waste Form for Hanford Waste Treatment Plant Secondary Waste - 12306

Hanford currently has 212,000 m{sup 3} (56 million gallons) of highly radioactive mixed waste stored in the Hanford tank farm. This waste will be processed to produce both high-level and low-level activity fractions, both of which are to be vitrified. Supplemental treatment options have been under evaluation for treating portions of the low-activity waste, as well as the liquid secondary waste from the low-activity waste vitrification process. One technology under consideration has been the THOR{sup R} fluidized bed steam reforming process offered by THOR Treatment Technologies, LLC (TTT). As a follow-on effort to TTT's 2008 pilot plant FBSR non-radioactive demonstration for treating low-activity waste and waste treatment plant secondary waste, TTT, in conjunction with Savannah River National Laboratory, has completed a bench scale evaluation of this same technology on a chemically adjusted radioactive surrogate of Hanford's waste treatment plant secondary waste stream. This test generated a granular product that was subsequently formed into monoliths, using a geo-polymer as the binding agent, that were subjected to compressibility testing, the Product Consistency Test and other leachability tests, and chemical composition analyses. This testing has demonstrated that the mineralized waste form, produced by co-processing waste with kaolin clay using the TTT process, is as durable as low-activity waste glass. Testing has shown the resulting monolith waste form is durable, leach resistant, and chemically stable, and has the added benefit of capturing and retaining the majority of Tc-99, I-129, and other target species at high levels. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Radioactive Waste Storage Facility at the Armenian NPP - 12462

We present a detailed contaminant transfer dynamics model for radionuclide in geosphere and biosphere medium. The model describes the transport of radionuclides using full equation for the processes of advection, diffusion, decay and sorption. The overall objective is to establish, from a post-closure radiological safety point of view, whether it is practical to convert an existing radioactive waste storage facility at Armenian NPP, to a waste disposal facility. The calculation includes: - Data sources for: the operational waste-source term; options for refurbishment and completion of the waste storage facility as a waste disposal facility; the site and its environs; - Development of an assessment context for the safety assessment, and identification of waste treatment options; - A description of the conceptual and mathematical models, and results calculated for the base case scenario relating to the release of contaminants via the groundwater pathway and also precipitation especially important for this site. The results of the calculations showed that the peak individual dose is < 7 E-8 Sv/y arising principally from I-129 after 700 years post closure. Other significant radionuclides, in terms of their contribution to the total dose are I-129, Tc-99 and in little C-14 (U- 234 and Po-210 are not relevant). The study does not explore all issues that might be expected to be presented in a safety case for a near surface disposal facility it mainly focuses on post- closure dose impacts. Most emphasis has been placed on the development of scenarios and conceptual models rather than the presentation and analyses of results and confidence building (only deterministic results are presented). The calculations suggest that, from a perspective the conversion of the waste-storage facility is feasible such that all the predicted doses are well below internationally recognized targets, as well as provisional Armenian regulatory objectives. This conclusion applies to the disposal of the ANPP present and future arising of L/ILW operating wastes. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Advancing Performance Assessment for Disposal of Depleted Uranium at Clive Utah - 12493

A Performance Assessment (PA) for disposal of depleted uranium (DU) waste has recently been completed for a potential disposal facility at Clive in northwestern Utah. For the purposes of this PA, 'DU waste' includes uranium oxides of all naturally-occurring isotopes, though depleted in U-235, varying quantities of other radionuclides introduced to the uranium enrichment process in the form of used nuclear reactor fuel (reactor returns), and decay products of all of these radionuclides. The PA will be used by the State of Utah to inform an approval decision for disposal of DU waste at the facility, and will be available to federal regulators as they revisit rulemaking for the disposal of DU. The specific performance objectives of the Clive DU PA relate to annual individual radiation dose within a 10,000-year performance period, groundwater concentrations of specific radionuclides within a 500-year compliance period, and site stability in the longer term. Fate and transport processes that underlie the PA model include radioactive decay and ingrowth, diffusion in gaseous and water phases, water advection in unsaturated and saturated zones, transport caused by plant and animal activity, cover naturalization, natural and anthropogenic erosion, and air dispersion. Fate and transport models were used to support the dose assessment and the evaluation of groundwater concentrations. Exposure assessment was based on site-specific scenarios, since the traditional human exposure scenarios suggested by DOE and NRC guidance are unrealistic for this site. Because the U-238 in DU waste reaches peak radioactivity (secular equilibrium) after 2 million years (My) following its separation, the PA must also evaluate the impact of climate change cycles, including the return of pluvial lakes such as Lake Bonneville. The first draft of the PA has been submitted to the State of Utah for review. The results of this preliminary analysis indicate that doses are very low for the site-specific receptors for the 10,000-year compliance period. This is primarily because DU waste is not highly radioactive within this time frame, the DU waste is assumed to be buried beneath zones exposed by erosion, groundwater concentrations of DU waste constituents do not exceed groundwater protection limits with in the 500-year compliance period, and the first deep lake occurrence will disperse DU waste across a large area, and will ultimately be covered by lake-derived sediment. A probabilistic PA model was constructed that considered DU waste and decay product doses to site-specific receptors for a 10,000-yr performance period, as well as deep-time effects. The quantitative results are summarized in Table VII. Doses (as TEDE) are always less than 5 mSv in a year, and doses to the offsite receptors are always much less than 0.25 mSv in a year. Groundwater concentrations of Tc-99 are always less than its GWPL except when the Tc-99 contaminated waste is disposed below grade. Even in this case, the median groundwater concentration is only 4.18 Bq/L (113 pCi/L), which is more than one order of magnitude less than the GWPL for Tc-99. The results overall suggest that there are disposal configurations that can be used to dispose of the proposed quantities of DU waste that are adequately protective of human health. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Current Status of Human Resource Training Program for Fostering RIBiomics Professionals

RI-Biomics is a state-of-the-art radiation fusion technology for evaluating in-vivo dynamics such as absorption, distribution, metabolism and excretion (ADME) of new drug candidates and biomaterials using radioisotope (RI), and quantitative evaluation of their efficacy via molecular imaging techniques and animal models. The RI-Biomics center is the sole comprehensive research and experiment complex in Korea that can simultaneously perform the radio-synthesis of drug candidate with radioisotope, analysis, and molecular imaging evaluation with animal model. Molecular imaging techniques, including nuclear imaging (SPECT and PET), near-infrared fluorescent (NIRF) imaging, and magnetic resonance imaging (MRI), are the cutting-edge technologies for evaluating drug candidates. Since they allow in vivo real-time imaging of the diseased site, monitoring the biodistribution of drug and determining the optimal therapeutic efficacy following treatments, we have integrated RI-ADME and molecular imaging to provide useful information for drug evaluation and to accelerate the development of new drugs and biomaterials. The RI-Biomics center was established with total investment of 18 million $ during four years from 2009 to 2012 in order to develop a comprehensive analyzing system using RI for new drug development as an axis for national growth in the next generation. The RI-Biomics center has labeling synthesis facility for the radiosynthesis of drug candidate with radioisotope such as Tc-99m, I-125, I-131, F-18, H-3 and C-14 using hot cell. It also includes RI-general analysis facilities, such as Radio-HPLC, LC/MS, GC/MS, gamma counter that can analyzing the radio-synthesized materials, and animal image analysis facilities that developed small animal imaging equipment such as SPECT/PET/CT, 7 T MRI, in-vivo optical imaging system and others. In order to achieve the system to verify safety and effectiveness of the new drugs using RI, it is necessary to establish a human resource training program for fostering RI-Biomics professionals in the following key fields; (1) Radio-pharmaceuticals synthesis and labeled compound development, (2) Development of RI-ADME in the living object and image assessment technology. Personnel training program that carries out theoretical education and practical training in the field related to RI-Biomics in parallel is being conducted. Internship training for university students has been administered twice already while educational program for the existing professionals in the RI-Biomics field will be carried out during the summer of 2014. The human resource training program for combination of RIADME and different molecular imaging techniques can offer synergistic advantages to facilitate understanding RIADME and fostering RI-ADME professionals. (authors)

62 RADIOLOGY AND NUCLEAR MEDICINE↗

Use of Test Beds to Facilitate Implementation of Innovative Technologies at Complex Sites - 17147

The Department of Energy, Office of Environmental Management (DOE-EM) has invested significant resources targeted at reducing risk to the environment and to the public. Considerable progress has been achieved, yet major remediation challenges remain. DOE-EM recognizes that new, innovative technologies offer the potential to improve operational performance, accelerate schedules, and reduce cost for environmental cleanup. One challenge that often limits the adoption of new technologies is the lack of information about their performance under real-world conditions. This is particularly true of the complex radiological and chemical environments present at many DOE-EM sites. Test-beds provide a useful approach to test new technologies and technical approaches. With the ability to be field-scale, meso-scale, and/or lab-scale in size and scope, test-beds allow scientists and researchers to explore theories, technologies and tools within replicated, 'real-world' physical, chemical and environmental conditions. To address this challenge, DOE-EM is managing a suite of test-beds and prototype test facilities. The focus of test-bed efforts is on several key issues and challenges that confront site operators. Priority environmental remediation concerns are technetium-99, cesium-137, and strontium-90. Additionally, DOE-EM has determined that advanced robotic and remote sensing techniques offer promise in increasing operational efficiency, reducing personnel exposure, and enhancing safety. Savannah River National Laboratory is implementing a test-bed program. Selection of technologies and alignment with test-beds will depend on applicability to DOE-EM issues, technology maturity, a site's ability to host vendors (industry, SBIR, academia, national laboratories), and availability of supporting infrastructure. Viable results are generated through the test-bed process so that these solutions can be utilized in larger-scale thereby facilitating quick implementation of new technologies at DOE sites. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Characterization and Well Remediation Strategies for Improved Long-term Pump and Treat System Injection Well Performance - 17272

In June of 2012 the CH2M Hill Plateau Remediation Company completed construction of a 2,500 gallon-per-minute (gpm) pump and treat (P and T) system designed to remove technetium-99, uranium, carbon tetrachloride, trichloroethene, nitrate, and total and hexavalent chromium from incoming groundwater. The entire pump and treat system consists of the following primary components: 1) extraction wells (20+); 2) a radionuclide removal facility; 3) a biological system; 4) sludge processing; 5) vapor processing; and 6) injection wells. Overall efficiency of the 200 West P and T has been adversely impacted due to fouling in vertical injection wells where treated water is being reinjected into the subsurface. Fouling of the wells is thought to be caused by bacteria producing extracellular polymeric substances (EPS), as well as high manganese and iron in the biofilm. Due to decreased specific injectivity, all injection wells are on a maintenance schedule requiring rehabilitation. During rehabilitation, the wells are out of service affecting the overall capacity of the P and T, plus the rehabilitation process adds hundreds of thousands of dollars of cost to operate the P and T each year. Chemical and biological foulants may be coming from the P and T or are already present in the subsurface and active pumping has stimulated growth of the microbes present. Chemical and microbiological characterization efforts are in progress for the biological plant and injection well system to determine causative agents in the biofouling. P and T effluent tank samples, injection well distribution tanks, bailed samples from the fouled well, and post-rehabilitation samples were analyzed. Chemical constituents analyzed include, sulfate, phosphorus, chemical oxygen demand, total organic carbon, and iron. Microbial characterization included total heterotrophs, anaerobic heterotrophs, nitrate reducing bacteria, iron reducing bacteria and sulfate reducing bacteria. Molecular biological tools employed included 16S rRNA meta-genomes, and functional gene analysis using qPCR. Chemical analysis of the P and T effluent and injection well tanks showed the presence of ∼6.5 mg/L nitrate, 96 mg/L sulfate and phosphorus levels below 1 mg/L. Bailed samples from the injection well showed similar concentrations for each chemical, but following cleaning, TOC, total iron, manganese and other constituents were higher than prior to cleaning. These data show that bacterial nutrients and metals that may precipitate are present in the well water and are likely the causative agent in the biofouling. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Sub-Grid Model Improving Shrinking Core Simulations of Redox-Sensitive Technetium Release From Slag Cement Waste Forms - 17295

Numerical simulation of Tc transport in reducing cementitious materials undergoing oxidation is challenging because the effective solid-solute partition coefficient (K{sub d}) varies greatly between the reduced and oxidized states, assuming the Tc mass loading far exceeds that required to reach solubility in the aqueous phase, as is typical in waste disposal. In a conventional finite-volume formulation, the local aqueous concentration increases orders of magnitude when a computational cell switches from a reducing to oxidizing condition, which is an artifact of numerical discretization. In principal such artifacts can be reduced to an insignificant level by refining the computational grid. However, this approach is not feasible in typical facility scale applications due to computer runtime and storage limitations. In this paper an alternative sub-grid Tc release model is derived to approximately capture redox processes occurring at scales smaller than the computational mesh. This alternative approach introduces minimal additional computational expense, while significantly reducing numerical artifacts in the form of Tc concentration / flux spikes. The method is illustrated for two simple geometries and the Savannah River Site Saltstone waste Disposal Facility. The sub-grid model is shown to significantly reduce artificial spikes in Tc-99 release, resulting in more physically realistic numerical simulations. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Removal of Technetium-99 on Ion Exchange Resin - A Case Study at 200 West Pump and Treat - 17303

The 200 West Pump and Treat (P and T), on the U.S. Department of Energy's Hanford Site and operated by CH2M HILL Plateau Remediation Company, has a 9,464 liters per minute capacity designed to capture and treat an 8 km{sup 2} contaminated groundwater plume and reduce the mass of contaminants of concern (COCs) by 95 percent in 25 years. The plume is a complex mix of contaminants including technetium-99 (Tc-99), uranium, carbon tetrachloride, chromium, hexavalent chromium, nitrate, and iodine-129 (I-129). Ion exchange (IX), a common treatment approach for Tc-99 used by many nuclear remediation sites, is used to remove uranium and Tc-99 at 200 West P and T. The IX vessels are arranged in a sequence of three to provide a lead, lag and polish vessel. Tc-99 concentrations are monitored monthly to assure good performance of the system. Performance tracking revealed surprising results. On October 14, 2015, the resin in the lead vessels released Tc-99 at concentrations greater than what was being fed. Although Tc-99 was captured by the downstream lag and polish vessels, the facility staff grabbled with the sudden release of Tc-99 entertaining a variety of scenarios that might result in sudden failure of the IX processes. Subsequent sampling indicated Tc-99 removal began a steady return to normal. The incident raised many questions, chief among them whether the same events could recur and undermine Tc-99 treatment. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗