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At least 991 records · Page 55
All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

Development of U and Pu Co-Recovery Process (Co-Processing) for Future Reprocessing

Co-processing process, which is the modified Purex process focused on co-recovery of Pu and U, has been studied at Operation Testing Laboratory, Tokai Reprocessing Plant in JAEA. The set up of the process was performed with flow-sheets study by process calculation to avoid Pu isolation in the whole process and to co-recover Pu/U product solution with a suitable Pu/U ratio (0.5< Pu/U <2). The initial Pu/U ratios of the feed solutions were taken as 1%, 3% and 20% considering the composition of the future spent fuels. The verification of the flow-sheets for each feed solutions were carried out with mixer-setters and active Pu/U feed solutions, focusing on the partitioning unit, and favorable back extraction performances of Pu accompanied by U were observed at all cases of the given feed solutions. According to these results, the co-processing process showed a good prospect to treat all kinds of future fuels from LWR, LWR-MOX and FBR, and a good prospect to be simplified by omitting the Pu/U purification cycle.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Fuel salt and container material studies for MOSART transforming system

A study is under progress to examine the feasibility of single stream Molten Salt Actinide Recycling and Transmuting system without and with Th support (MOSART) fuelled with different compositions of actinide tri-fluorides (AnF{sub 3}) from used LWR fuel. New fast-spectrum design options with homogeneous core and fuel salts with high enough solubility for AnF{sub 3} are being examined because of new goals. The flexibility of single fluid MOSART concept with Th support is underlined, particularly, possibility of its operation in self-sustainable mode (Conversion Ratio: CR=1) using different loadings and make up. The paper summarizes the most current status of fuel salt and container material data for the MOSART concept received within ISTC-3749 and ROSATOM-MARS projects. Key physical and chemical properties of various fluoride fuel salts are reported. The issues like salt purification, the electroreduction of U(IV) to U(III) in LiF-ThF{sub 4} and the electroreduction of Yb(III) to Yb(II) in LiF-NaF are detailed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Optical spectra and luminescence of REE and TRU for analytical purposes in chloride melts

This paper analyzes absorption spectra of molten salts containing some of the actinide and rare-earth elements (REE) and evaluated the prospects of using the individual transitions to control the composition of the spent molten salts using adsorption and luminescence spectroscopy from the standpoint of the theory of Judd-Ofelt. It is shown that the fluorescent method can be recommended only for the monitoring of the degree of purification of the molten salt from the REE in the final stage, when only trace amounts of fission products are in the molten salt. In this case, the content of REE in the molten process is much smaller than that of the used model samples and that eliminates the problem of a significant decrease in the population levels of fluorescent Eu{sup 3+} by quenching impurities. In addition, the presence of salt in the matrix of americium and curium can cause self-luminescence of Eu{sup 3+} due to alpha disintegration energy of transuranics, which also complicates the quantitative measurement of luminescence for the determination of REE.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

PAREX, A numerical Code for process design and integration

The modeling approach for process design is illustrated in this paper by the work done to qualify the PAREX code to design the flow sheet of a uranium/plutonium splitting operation in a pulsed column. The different flow sheet experimentations are presented along with the simulation results. The qualified code has been successfully used to design the flow sheet of this operation in the first purification cycle of UP2-800 plant at La Hague. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Shipping Grossly Dewatered resins in the RT-100 Type B Transportation Cask - 16068

Ion-exchange resins are used in the purification of coolant water in boiling water reactor nuclear power plants. The spent resins are a form of low-level or intermediate-level radioactive waste. These resins may have a relatively high specific activity level and are normally transported to a disposal facility within a secondary container placed in a Type B transportation cask. During transportation, hydrogen gas may build up in the cask due to radiolysis of the hydrogenous material in the contents. In the licensing process, a Type B cask is carefully analyzed to ensure that the maximum amount of hydrogen gas generated does not exceed 5%. The calculation to determine the mole fraction of hydrogen gas generated is a function of many variables, including the properties of the contents and the volume ratios of the contents. Because water generates hydrogen gas when it undergoes radiolysis, the amount of water present in the waste is normally limited. Typically, Type B transportation casks are allowed to carry dewatered resins, meaning less than 1% free-standing water by volume. To ensure compliance with this limitation, the dewatering process requires specialized equipment to pump the water from the secondary container. In an effort to enhance the usability of the RT-100 Type B transportation cask, Robatel Technologies re-analyzed the hydrogen gas generation calculation for contents including 'grossly dewatered' resins. Grossly dewatered resins are defined as resins containing less than 20% free-standing water by volume. This higher limit is made possible by a more thorough hydrogen gas generation calculation, which ensures that the flammability limit is not reached during shipment. In July 2015, The USNRC issued Revision 1 of the Certificate of Compliance for the Robatel Technologies RT-100 Type B transportation cask. This revision allows the cask to be used for grossly dewatered resins. With the revised Certificate of Compliance, the RT-100 is the first of its kind whose authorized contents include greater than 1% free liquids by volume. As a result of this change, the cask user can meet the 20% maximum free-standing water volume with a much simpler gravity-driven dewatering process. It is no longer necessary to have a complex dewatering system to ensure compliance with regulatory requirements. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

WOW Technology's Innovative Radioactive Liquid Waste Treatment - 16128

WOW presents its revolutionary technology and equipment for liquid radioactive waste treatment: outperforming ultimate water decontamination and purification process, enhanced sludge concentration, no secondary waste nor consumables, fully automated, remote controlled and self-decontaminating device. The WOW's technology is based upon a never before observed discovery of fluid dynamics science: the possibility of performing a molecular separation between solute and suspended elements and the solvent. The combination of such a molecular separation process with a standard vacuum evaporation improves the abatement performances by thousands of times, with respect to those of the state of the art vacuum evaporators. In addition to this, no secondary waste is produced during the process, as no filters, membranes, resins or additives are used. WOW equipment, automated and remote controlled, self decontaminates after use and can be designed and constructed either tailored to the application needs or with a modular approach for enhanced transportability and application flexibility. After the preliminary verification by CNR, the Italian National Research Center, WOW Technology's decontamination device was tested c/o LENA, the Laboratory of Applied Nuclear Energy of the University of Pavia, Italy with a simulated solution 6000 times more contaminated than the nuclear reactor's cooling water of Fukushima-Daiichi NPP. In addition to that, WOW's technology was also used in a real case at the Radiochemistry laboratory of the Pavia's University Chemistry department. Both the above mentioned contaminated fluids have been successfully decontaminated without production of additional or secondary waste. WOW's technology has already performed on industrial scale c/o the Nuclear Repository of S.S.M. in Saluggia, Italy: 45000 liters of acid radioactive solution have been successfully decontaminated to a Decontamination Factor (DF) of 335000 for Cs-137 by one single evaporation step and without using any additional tool to control steam. [Performances independently certified by: National Physics Laboratory (UK) and Pavia University Radiochemistry Laboratory, Chemistry Department (Italy)]. The high decontamination factors and the high volume reduction of the sludge make this technology suitable for the ultimate treatment of highly contaminated liquid waste such as the Fukushima-Daiichi nuclear reactor's cooling water. The WOW's technology addresses as well to interventions in both, operating and decommissioning nuclear power plants and any other nuclear and radioactive facility whose produced and/or stored radioactive liquid solutions, including acid liquid solutions, can be successfully treated, the waste volume reduced and no secondary waste produced. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Radiochemical Method for Characterization of the Filter Cartridges from the IEA-R1 Reactor - 16324

The filter cartridges used in water purification system of the IEA-R1 research reactor become radioactive waste after the end of their useful life. The characterization of this waste is one step in their management, which aims at identifying and quantifying the radionuclides present, including those known as 'difficult to measure' (DTM) radionuclides. Wastes from nuclear reactors contain fission and activation products and transuranic elements, only few of them emitting gamma radiation measurable by simple gamma spectrometry methods. In routine waste management activities, the concentrations of the DTM can only be estimated by indirect methods such as scaling factors. The method of the scaling factors uses empirically determined proportions between the concentrations of DTM radionuclides and that of easily measurable gamma emitters called key nuclides (KN), to calculate DTM concentrations. Determining the scaling factors for the filter cartridges is a hard task when the number of samples is large, not only because the preparation of samples is difficult but also because the number of radionuclides to analyze is large. If it can be demonstrated that DTM and KN are both distributed evenly in the filters, less samples from each filter are necessary, reducing lab work. The aim of this paper is to present the results of the assessment of homogeneity along the filter cartridges by gamma spectrometry of chemically reduced samples. Five thin slices of five filters were burnt and the ashes quantitatively dissolved and the activity concentration of the solutions determined by gamma spectrometry. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Environmental Characterization Scans Using Sodium-Iodide Gamma Spectroscopy to Determine Ra-226 Concentration in Surface Soil - 16326

The recent natural gas drilling boom has brought heightened attention to the issue of TENORM waste management. Radioactive residues from the extraction, treatment, and purification of materials containing uranium and thorium, or their progeny, pose a public health hazard primarily in the form of ground water contamination. Hydraulic fracturing can liberate formation waters from shale; the high salinity of which can cause radium to move with the water to the surface. The disposal and handling of this water has been a subject of public controversy. For soil remediation projects, radium-226 is generally one of the radionuclides of concern due to its alpha emissions. However performing a survey scan for Ra-226 is difficult because its photon emissions are low energy and have minimal probability of emission per decay (186.2 keV, 3.3% abundance). Additionally resolving its emission from U-235 (185.7 keV, 57% abundance) with field instrumentation is not feasible. Therefore detection of Ra-226 relies on the presence and photon emissions of its progeny. This paper outlines methodology for calculating the minimum detectable concentration of Ra-226 in characterization scans using a sodium-iodide detector array coupled with a multi-channel analyzer. Using progeny ingrowth as a detection proxy requires knowing decay time in order to determine progeny emission rates. Knowledge of site history, process, and prior survey results can be key in determining an accurate decay time. However in the absence of such information in-situ measurements can be used to estimate decay time by comparing observed progeny emission ratios against established values. These values are determined using Oak Ridge National Laboratory's ORIGEN-ARP program to generate photon energy spectra for Ra-226 and its progeny at specified time intervals. Once decay time is established, the detector's minimum detectable concentration can be calculated by synthesizing the expected emission spectra and Los Alamos National Laboratory's Monte Carlo N-Particle transport code to simulate detector response. Traditionally, one of the steps in calculating a system's minimum detectable concentration is to place an appropriate button source under the detector, record the response, and then move the source by a specified increment. This process is repeated until the detector's spatial efficiency has been mapped. Due to the impracticality of performing this with Ra-226 button sources of varying decay times, Monte Carlo N-Particle transport code was utilized to determine the detector's spatial efficiency. A dime sized source with the established emission spectra was placed into the center of a rectangle of soil. This rectangle, including the source, was used as a lattice element to fill the region of space below the detector. The simulation was designed to track the energy distribution of pulses created in the sodium iodide crystal using the F8 tally, and was setup to isolate each source; thereby correlating detector response to source position. The simulation output required post-processing to determine conversion factors from counts per second to disintegrations per second. Post-processing accounted for subtraction of the Compton continuum, and average the data over the scanning pattern to obtain the expected spatial sensitivity during a characterization study. This was performed for the following photo-peaks: 609 keV (Bi-214), and 1764 keV (Bi-214). For calculation of the scan minimum detectable concentration, the surveyor efficiency coefficient from NUREG 1507 was supplanted in favor of a scan speed deviation factor. This was used because the detector utilizes computerized data collection (spatial and spectral). This report finds that, for two shielded 10.16 cm x 10.16 cm x 40.64 cm sodium-iodide crystals with integrated multi-channel analyzers, the minimum detectable concentration in surface soils is less than 1 pCi per gram Ra-226 for a 1 second survey measurement. These results assume a volumetric source with a maximum depth of 15.24 cm and a minimum decay time of 21 days. Verification of the system's capabilities was performed by sending soil samples from scanned areas, and locations where in-situ measurements were performed, to an independent laboratory for analysis. The laboratory results demonstrated that scanned area was principally comprised of soil containing less than 1 pCi per gram Ra-226. This borders the system MDC and it underestimated the concentration of radium in the soil. The probable causes of this is escape of radon gas from the surface soil. Radon is the daughter of radium, and has a high mobility in soil. If the gas escapes, then the progeny used as the detection proxy will be present in lower than expected concentrations. Radionuclide identification using high volume sodium-iodide crystals in low level characterization scans is demonstrated to be possible. Further system refinement will focus on developing a methodology to estimate a radon escape factor for site specific use. Accounting for radon escape in the top 15.24 cm of soil will allow the system to better estimate Ra-226 concentrations in soil. This is a significant advancement over the gross counting statistics used in most studies to satisfy MARSSIM survey requirements, because it will increase stakeholder confidence. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Daher-TLI UCSC Wash Process - Lessons Learned from Handling a Legacy Cylinder - 16506

The Daher-TLI UF{sub 6} Cylinder Service Center (UCSC) began operation in 2013 and functions as a dedicated wash, re-certification, storage and disposal facility for 122 cm uranium hexafluoride (UF{sub 6}) cylinders. The primary purpose of the facility is to provide cylinder cleaning and re--certification services to the UF{sub 6} industry for cylinders that previously contained natural or depleted UF6. The facility also has the ability to recover most of the uranium in the form of uranyl peroxide (UO{sub 4}) or triuranium octoxide (U{sub 3}O{sub 8}). Small residues of UF{sub 6} (commonly referred to as a heel) from natural cylinders are removed and treated in a patented three-step purification process that recovers U{sub 3}O{sub 8} that meets the ASTM specification. The wastewater ordinarily contains a mixture of sodium salts (77% sodium sulfate, 23% sodium nitrate) along with minute traces of uranium (1 - 2 mg/L) and fluoride (10 - 20 mg/L). Over a period of 6 months, lab analyses were performed on the filtrate from a legacy cylinder in an attempt to separate uranium and fluoride from this stream. These pilot results established near quantitative recovery of uranium (< 1 mg/L) and near complete separation of fluorides (5 - 15 mg/L). This paper follows the efforts of the Daher-TLI UCSC to process the abnormal heeled UF{sub 6} cylinder. This paper also documents the lessons learned from this experience and how those lessons were used to improve the UCSC wash process. These modifications served to recover 99.87% of the input uranium as UO{sub 4} while simultaneously creating a waste stream with only trace amounts of impurities. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Decontamination and Decommissioning Activities of a Radiopharmaceuticals Production Facility - 18067

On May 2012, Best Medical Belgium S.A. (BMB), a manufacturer of pharmaceutical products located on the nuclear site of Fleurus, was declared bankrupt. According to the Belgian law, ONDRAF/NIRAS (Belgian Radioactive Waste Management Agency) was entrusted with the clean-up and decommissioning operations of the former BMB facilities, nowadays known as 'ONDRAF Site Fleurus' (ONSF). In June 2015, ONSF launched a Decontamination and Decommissioning (D and D) program, for which Tractebel: - Supports the definition of the clean-up and decommissioning strategies; - Drafts the Final Decommissioning Plan (FDP) and Safety Analysis Report (SAR) that will be submitted to FANC/AFCN (Belgian Federal Agency for Nuclear Control) in order to get the License for decommissioning; - Drafts the technical specifications of the public tendering process related to the D and D program; - Ensures the day to-day follow-up of the D and D activities on site. Four nuclear buildings are part of the D and D activities. One of them, building 14 (B14), represents the most challenging and extensive task with the clean-up of highly contaminated areas and its further decommissioning up to the final state. This building features two cyclotrons (CGR and IBA cyclotrons) successively used to produce radioisotopes, as well as hot cells and glove boxes used for their purification and the conditioning of the final pharmaceutical product. The present paper aims at presenting the most recent development steps of the clean-up phase regarding the adaptations of existing facilities and the upcoming activities in preparation of the decommissioning activities. To allow the sorting and the treatment of the radioactive material identified in the physical and radiological inventory, several workshops have been implemented within the B14 building in order to deal with radioactive materials arising from the clean-up and D and D activities. Following the conceptual design phase, Tractebel has actively monitored the on-site activities for the construction of those workshops such as: - The modification of the civil works; - The revamping of the ventilation system; - The installation of dedicated lifting equipment and workstations equipped with cutting devices; - The assembly of a docking station for filling the radioactive packages. The removal of the activated concrete surrounding the IBA cyclotron requires the erection of a containment structure and an extension structure. The containment structure, composed of light metallic beams, is located above the cyclotron and its irradiation zones. For operational security reasons and working efficiency, the cutting process of the concrete ceilings and walls will be performed from the top inside the containment structure by means of an eight tons overhead bridge crane. Concrete blocks resulting from the cutting process will then be lifted, without any loss of containment, towards the extension zone of the nuclear zone of the IBA cyclotron for further resizing, sorting and evacuation towards the identified waste routes. To prepare this task which will be performed during the decommissioning phase, Tractebel sets up the necessary technical specifications. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Radiochemical Characterization for Decommissioning Purposes at NRG: The Case of Ion Exchange Resins - 18184

The High Flux Reactor (HFR) in Petten (NL) has been producing medical isotopes for several decades. Additionally, the HFR is used for many experiments including fuel/material tests. Over the years, Ion Exchange Resins (IER) that are used for the purification of the primary circuit and the basin have been replaced. Spent IER were accumulated and stored on site. Over time, the IER has become a historical waste issue that needs to be solved pragmatically as space to store them on site becomes scarce. This paper provides an overview of the characterization methodology needed to manage the Ion Exchange Resin wastes. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Electrochemical behavior of Nd and its co-reduction with Bi in molten LiCl-KCl salts - 18360

A molten salt phase is used as the solvent in pyrochemical reprocessing. The molten salt waste generated in pyro-processes contains the residual lanthanides and actinides as various compounds. Various methods have been proposed for recovering and recycling these useful elements. The separation of actinides from lanthanides is performed by electrolysis or reductive extraction. Herein we have studied PyroRedSox-based co-reduction of Nd and Bi via electrolysis in a molten chloride salt. The redox behavior of Nd and the formation potential of intermetallic compounds of Nd and Bi were determined using cyclic voltammetry (CV). Electrorefining was performed under a constant reductive potential at 773 K in LiCl-KCl eutectic mixture; the co-reduction potential was -1.37 V. Lastly, the co-reduction potential of the intermetallic compound was applied to reduce the amount of Nd and Bi. Scanning electron microscopy (SEM) revealed that various kinds of intermetallic compounds were formed in Bi and Nd, and were deposed in the lower part of the molten salt, allowing purification of the contaminated molten salt. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

CNL Technologies for Heavy Water Detritiation - 19437

Canadian Nuclear Laboratories (CNL) is responsible for a large amount of heavy water owned by the Government of Canada. A significant portion of this heavy water has been downgraded and is chemically and/or radiologically contaminated (besides tritium). Storage, monitoring and disposal of this water incurs significant costs. Upgrading following chemical purification (including removal of radionuclides other than tritium) of this water is feasible with commercial processes. However, removal of the tritium to the level that would allow the heavy water to be available for non-nuclear applications requires very specialized processes. CNL has developed and demonstrated the Combined Electrolysis and Catalytic Exchange (CECE) detritiation technology to reduce the tritium to the level at which the heavy water can be classified as non-radioactive. The CECE process consists of a Liquid Phase Catalytic Exchange (LPCE) bed, where the transfer of hydrogen isotopes between water and gaseous hydrogen occurs, and an electrolysis cell (E-cell) that produces hydrogen and separates the isotopes of hydrogen, adding to the enrichment achieved in the LPCE catalyst bed. The application of the CECE technology to the tritium-contaminated heavy water would restore its full market value. As part of a project supported by CNL Decommissioning and Waste Management, a business case is being developed to build and operate a facility that will purify and detritiate heavy water to render it commercially attractive. The main goal of the current study is to evaluate the net benefits of detritiation and product sale versus the cost of permanent storage or disposal. Such a facility would not only eliminate the need to store the heavy water as a radioactive waste material and provide revenue, but would also serve as a demonstration of CNL's technology for similar applications elsewhere. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Water and Ammonia Desorption from 5A Zeolite

The objective of this project is to demonstrate the adsorption and desorption characteristics of water and ammonia on 5A zeolite during regeneration conditions. As discussed in literature, zeolite material is a crystalline hydrated aluminosilicate material. Zeolites have high cation exchange ability and are commonly used as adsorbents in separation and purification processes. The property of uniformly sized pores allows zeolites to selectively separate gases. Because they have varying ion selectivity, there is competitive adsorption in a multicomponent mixture. 5A zeolite has a uniform pore diameter of 5 Angstroms. Water and ammonia are miscible components with similar molecular weights and size. This makes distinguishing the two compounds difficult using mass spectrometry. In a previous study, it was shown that water tends to desorb from 5A zeolite from 40 deg. C to 353 deg. C with peaks at 177 deg. C and 244 deg. C, while ammonia tends to desorb from 300 deg. C to 580 deg. C with a peak at 436 deg. C as determined using a Micromeritics AutoChem 2920 Instrument. The one drawback of using a Micromeritics Instrument is that only a few milligrams of sample material can be used. Bench scale tests can better demonstrate desorption characteristics of water and ammonia during regeneration from 5A zeolite material. Water in argon carrier gas and ammonia in hydrogen carrier gas will be tested separately, then a mixture of the two will be tested. Initial bake-out conditions: hydrogen carrier gas, slowly raise temp to approximately 450 deg. C until desired moisture content <100 ppm is achieved. Water loading: argon carrier gas through bubbler until breach is observed using the residual gas analyzer (RGA). Ammonia loading: hydrogen carrier gas through system until breach is observed using RGA. Regeneration conditions: hydrogen carrier gas, raise temp to approximately 300 deg. C, then slowly to 450 deg. C. Observe signals on RGA. Weigh samples between tests and integrate RGA signals. Accomplishments: Developed work control documentation and completed safety related requirements; Flow sheet and test system development; Initial bake-out of first sample. Testing will determine if regenerating zeolite material can be safely accomplished by high-temperature bake-out to remove ammonia. If unsuccessful, ammonia would only be removed at the end of life.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

DEM and MELT In-Can Vitrification Process for Fukushima Daiichi Water Treatment Secondary Waste - 20034

Fukushima Daiichi Nuclear Power Station was severely damaged by a tsunami induced by the earthquake on March 11, 2011. A large amount of water treatment secondary waste has been generated from the purification work of contaminated water that consists of the water flowing into the reactor building. The Mid and Long-Term Road map specifies that solid waste at Fukushima Daiichi Nuclear Power Station should be stored safely and reasonably according to its characteristics. In order to meet this goal, a versatile treatment solution is needed, and the In-Can vitrification seems to be a suitable solution. The DEM and MELT In-Can vitrification process is being developed and qualified in France to treat waste arising from decommissioning and dismantling operations. This qualification is carried out through the consortium of the DEM and MELT project coordinated by CEA which gathers Orano, ECM technologies and Andra. The DEM and MELT In-Can vitrification process is a robust, simple and versatile in situ vitrification process. It is designed for high or intermediate level waste and is compact enough to be implemented in an existing facility or close to the waste to be treated. It is developed to treat liquid and solid waste, to produce a small amount of secondary waste and to minimize investment and operating costs. The DEM and MELT In-Can vitrification process is also developed with a modular design in order to be adapted to nuclear operators needs and requirements. The In-Can melting process has already been developed and fully qualified by the CEA for Alpha liquid waste and can be rated TRL 7 for this scope. The In-Can vitrification technology is considered as a qualified solution that could be applied for the treatment of waste resulting from the D and D work in Fukushima Daiichi Nuclear Power Station. The feasibility evaluation of the DEM and MELT In-Can vitrification process for Fukushima Daiichi water treatment secondary waste is a project led by Orano, CEA and ANADEC. This R and D project is performed by the funds from the Ministry of Economy, Trade and Industry as The Subsidy Program 'Project of Decommissioning and Contaminated Water Management'. This paper presents the first results of this feasibility evaluation, from the scenarios and glass formulation definition to the industrial implementation feasibility. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Analysis of Using a Hybrid ZIRCEX Process to Generate High-Assay Low-Enriched Uranium (HALEU) - 20158

Idaho National Lab (INL) is investigating the feasibility of providing High-Assay Low Enriched Uranium (HALEU) by recovering Highly Enriched Uranium (HEU) from spent nuclear fuel (SNF) and downblending the HEU with natural, low-enriched, or depleted uranium (NU, LEU, or DU). That investigation identified that significant quantities of HALEU could be produced with a Hybrid ZIRCEX process to further advanced reactor research while reducing INL's environmental liability. Many new advanced reactor designs are being developed with improved safety, efficiency, and economics. Most require nuclear fuel with U-235 enrichment between 5% and 20%, which is defined as HALEU. Although technically feasible, there is no current domestic capability to make HALEU in the US. While it is anticipated that industry will provide HALEU through commercial enrichment once advanced reactors mature, until a market is established, an interim source of HALEU is needed to enable research and demonstration. An interim source of HALEU could be provided by recovering the HEU in some DOE-managed SNF and downblending it to between 5% and 20% using the Hybrid ZIRCEX Process. ZIRCEX is a dry head-end process to remove cladding (zirconium or aluminum) from SNF. Drying of the SNF is needed prior to introduction to ZIRCEX. After cladding removal, uranium and fission products in the bed material are oxidized and elutriated. In a Hybrid ZIRCEX process, the HEU process stream is sent to a very compact, modular solvent extraction process for uranium purification. The uranyl nitrate from solvent extraction is denitrated and calcined to produce a HALEU product ready for fuel fabrication. The long-lived fission products from solvent extraction are immobilized in glass using a small in-can melt. By removing the cladding before dissolution, the volume of glass waste is reduced by a factor of up to 300 times (when compared to not removing the cladding from the fuel). Other radioactive wastes from the process could be disposed of as Class A or Class B low-level waste. In addition to producing HALEU for advanced reactors, consumption of SNF would reduce DoE's environmental liability. Incorporating low-enriched spent fuel (versus exclusive use of LEU, DU, or NU) into the downblending scheme could reduce the SNF inventory even further. Several downblending and enrichment scenarios using INL fuels were investigated in this study. The results of this analysis show that the Hybrid ZIRCEX process could make a significant contribution to needed HALEU feedstock and reduce the environmental liability of managing SNF at INL. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Contaminant Migration Modeling to Support the In Situ Decommissioning of Hardened Nuclear Facilities - 20215

The decommissioning of hardened nuclear facilities provides a unique challenge in balancing current worker dose with exposure to future receptors and media. In situ decommissioning (ISD) is a cost-effective and safe option for the closure of such facilities but requires a detailed understanding of the threat that radioactive and hazardous constituents remaining in the facilities pose to groundwater and surrounding surface water. The Savannah River Site (SRS) extensively employs contaminant migration models within the decommissioning process to develop site-specific removal, grouting, and monitoring strategies in support of safe and effective final end states of nuclear facilities. This work discusses the use of contaminant migration modeling in the ISD process at SRS, including specific examples from the closure of P-Reactor and R-Reactor, Building 235-F, F-Canyon and F/H Laboratory Complex. Migration models developed during the closure of P-Reactor and R-Reactor were divided into four source areas: the reactor vessel, the process area, the disassembly basin, and the purification wing. Each source area was assigned a specific inventory and migration pathway and was then subject to varying hypothetical removal, capping, and grouting schemes. Modeling indicated that groundwater Maximum Concentration Limits (MCLs) might be exceeded if no action was taken for eleven and ten constituents of concern (COCs) at P-Reactor and R-Reactor, respectively. These exceedances could occur in as few as 200 to 500 years. Alternatively, the migration modeling demonstrated that the selected ISD actions reduced contaminant mobility which allowed for significant radioactive decay and resulted in fewer predicted exceedances of groundwater MCLs (five COCs for P-Reactor; eight COCs for R-Reactor). In response to the modeling results for P-Reactor and R-Reactor, effectiveness monitoring programs were developed to target contaminants, specific to each reactor, that may migrate to groundwater. Contaminant migration modeling also revealed that roof collapse was a large factor in the release of COCs to the environment, giving rise to roof improvements, and an inspection and vegetation control program to ensure roof stability over time. Contaminant migration modeling is also aiding in the closure planning for hardened facilities in F Area. Building 235-F housed the Actinide Billet Line, which produced Np-237 billets for irradiation in SRS reactors, and the Plutonium Fuel Form (PuFF) facility that produced Pu-238 heat sources for the space program. As a result of these missions, areas within Building 235-F contain considerable residual amounts of both Pu-238 and Np-237. Contaminant migration modeling of Building 235-F was originally performed in 2012 to identify the feasibility of ISD and the amount of radioactive material removal required to prevent the exceedance of groundwater MCLs. The original model indicated that a 60% reduction in the PuFF facility Pu-238 inventory could keep groundwater concentrations below standards, while Np-237 did not pose a threat to groundwater. However, updates to the model with an emphasis on source impact pathways revealed that, due to the orientation of the source areas relative to groundwater flow, no amount of reasonable removal of Pu-238 would keep groundwater concentrations below MCLs and that Np-237 could be a large contributor to localized MCL exceedances. With this insight, the refined 2019 model is being used to assist in the development of grouting plans specific to each facility source area, where bentonite may be utilized to slow the migration of Pu-238 and its daughter products from the PuFF facility and a reducing grout may decrease Np-237 transport by ensuring the nuclide remains in the less mobile +IV oxidation state. The beginning phases of contaminant migration modeling are underway for F-Canyon and associated facilities using lessons learned from the Reactors and Building 235-F. A contaminant migration pathway, similar to the pathway used for the reactor vessels, is being developed for the hot and warm canyons. The F-Canyon inventories are being spatially refined to identify specific and localized source zones, comparable to Building 235-F, that may impact groundwater. Contaminant migration modeling provides vital input within the ISD process, from facility investigation to post closure effectiveness monitoring, making it a valuable tool in the development of safe and effective end states for hardened nuclear facilities at SRS. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

A Higher Fidelity Cost Analysis of Wind and Uranium from Seawater Acquisition symbiotic Infrastructure

The vastness of the worlds oceans makes the low concentration of naturally existing uranium, 3.3 ppb, appealing as a recoverable natural resource. Although ample uranium is available currently through conventional means, uranium from seawater is a widely studied topic due to the long term security it can assure the nuclear fuel cycle. It is important to mention that seawater uranium is not meant to act as a direct competitor to terrestrial uranium, but instead create a cost ceiling that provides economic stability and potential savings. Additionally, seawater uranium could circumvent some of the environmental impacts associated with recovery of any land based resource. The system currently studied by a nation-wide consortium of national lab and university partners involves the passive recovery of uranium using polymer based adsorbents. High Density Polyethylene fibers undergo a radiation induced graft co-polymerization process involving amidoxime, to attract uranium, and a polar co-monomer to increase hydrophilicity. After further chemical conditioning adsorbent fibers are braided into 60 meter strands for marine deployment. Braids are moored to the ocean floor for the duration of their soaking campaign. After sufficient seawater exposure they are winched up so the adsorbed uranium may be eluted off the braids. This deployment and elution process is repeated multiple times before the adsorbent's ultimate disposal, where its lifetime is dictated by the degradation it suffers with each re-cycle. Previous economic analyses have identified the adsorbent production and mooring as the most expensive components of the recovery process. Therefore, a later publication by Picard et al. proposed an alternative deployment method in an effort to reduce seawater uranium production cost. The Wind and Uranium from Seawater Acquisition symbiotic Infrastructure (WUSABI) couples uranium recovery with off shore wind to reduce mooring capital cost. The uranium harvesting structure attached to the underwater base of the wind turbine supports elution tanks and an adsorbent pulley system. This allows for a nearly autonomous mooring and elution procedure as adsorbent is in constant motion, reaching the elution tanks at the end of its soaking campaign. Although the original publication did include a short economic section, further analysis was carried out to provide a higher fidelity production cost estimate for uranium recovered by this system. Sufficient detail was provided regarding the capital and operating costs of the novel uranium harvesting units. The adsorbent production and uranium elution and purification steps however were estimated by simpler methods, for example applying a previously reported cost for producing a unit mass of adsorbent. This methodology, although satisfactory for the zeroth-order approximation desired in the original publication, overlooks many of the complexities and feedbacks embedded in the full cost model, which are necessary to provide a more accurate estimate. Perhaps more importantly, given the constant updates regarding adsorbent synthesis and performance this value quickly become outdated. Syncing an improved estimate for the WUSABI harvesting units with the existing economic analysis for adsorbent production and elution, which evolves in tandem with the technology, provides a more robust production cost estimate that will remain current over time. In addition to the updates and higher level of detail provided in this independent analysis, consistency in methodology allows for accurate comparison of uranium production costs across recovery methods. (authors)

07 ISOTOPE AND RADIATION SOURCES↗