Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “reprocessed”

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 343 records · Page 19
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Modeling and validating tritium transfer in a grassland ecosystem in response to {sup 3}H releases

In this paper a radioecological model (TOCATTA) for tritium transfer in a grassland ecosystem developed on an hourly time-step basis is proposed and compared with the first data set obtained in the vicinity of the AREVA-NC reprocessing plant of La Hague (France). The TOCATTA model aims at simulating dynamics of tritium transfer in agricultural soil and plant ecosystems exposed to time-varying HTO concentrations in air water vapour and possibly in irrigation and rain water. In the present study, gaseous releases of tritium from the AREVA NC nuclear reprocessing plant in normal operation can be intense and intermittent over a period of less than 24 hours. A first comparison of the model predictions with the field data has shown that TOCATTA should be improved in terms of kinetics of tritium transfer.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Iodine Adsorption by Ag-Aerogel under Prototypical Vessel Off-Gas Conditions

U.S. regulations will require the removal of 129 I from the off-gas streams of any used nuclear fuel (UNF) reprocessing plant prior to discharge of the off-gas to the environment. The required plant decontamination factor for iodine will vary based on fuel burnup, cooling time, and other factors but is very likely to be >1000 and could be as high as 8000. Multiple off-gas streams within a UNF reprocessing plant combine prior to environmental release, and each of these streams contains some amount of iodine. To achieve the decontamination factors (DFs) that are likely to be required by regulations, iodine removal from the vessel off-gas will be necessary. The vessel off-gas contains iodine at very dilute concentrations (ppb levels), and will also contain water vapor. Iodine species present are likely to include both elemental and organic iodides. There will also be solvent vapors and volatile radiolysis products. The United States has considered the use of silver-based sorbents for removal of iodine from UNF off-gas streams, but little is known about the behavior of those sorbents at very dilute iodine concentrations. The purpose of this study was to expose silver-functionalized silica aerogel (AgAerogel) to a prototypical vessel off-gas stream containing 40 ppb methyl iodide to obtain information about organic iodine capture by silver-sorbents at very low iodine concentrations. The design of this extended duration testing was such that information about the rate of adsorption, the penetration of the iodine species, and the overall system DF could be obtained. Results show that CH 3 I penetrates into a AgAerogel sorbent bed to a depth of 3.9 cm under prototypical vessel off-gas conditions. An iodine loading of 22 mg I/g AgAerogel was observed in the first 0.3 cm of the bed. Of the iodine delivered to the system, 48% could not be accounted for, and future efforts will investigate this concern. Direct calculation of the decontamination factor is not possible, as no iodine was observed to break through the sorbent beds. Continued studies on the adsorption of iodine from prototypical vessel off-gas streams by silver-based sorbents will attempt to resolve some of the questions raised here, both regarding mass balance and the effect of aging on iodine adsorption by AgAerogel from a dilute gas stream. Additionally, the adsorption of different iodine species, such as I 2 and C 12 H 25 I will be studied. Other variables that merit examination are the gas velocity of the test and the dependence of the observed results on the inlet iodine concentration. Finally, longer duration testing should be considered in an effort to determine the mass transfer zone associated with iodine adsorption by AgAerogel under prototypical vessel off-gas conditions. The estimation of mass transfer zone is required for any future industrial implementation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Facility Modeling Capability Demonstration Summary Report

A joint effort has been initiated by Los Alamos National Laboratory (LANL), Oak Ridge National Laboratory (ORNL), Savanah River National Laboratory (SRNL), Pacific Northwest National Laboratory (PNNL), sponsored by the National Nuclear Security Administration’s (NNSA’s) office of Proliferation Detection, to develop and validate a flexible framework for simulating effluents and emissions from spent fuel reprocessing facilities. These effluents and emissions can be measured by various on-site and/or off-site means, and then the inverse problem can ideally be solved through modeling and simulation to estimate characteristics of facility operation such as the nuclear material production rate. The flexible framework called Facility Modeling Toolkit focused on the forward modeling of PUREX reprocessing facility operating conditions from fuel storage and chopping to effluent and emission measurements.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Evaluation of Ruthenium Capture Methods for Tritium Pretreatment Off-Gas Streams

In the reprocessing of used nuclear fuel, radioactive elements are released into various plant off-gas streams. While much research and development has focused on the abatement of the volatile nuclides 3 H, 14 C, 85 Kr, and 129 I, the potential release of semivolatile isotopes that could also report to the off-gas streams in a reprocessing facility has been examined. Ruthenium (as 106 Ru) has been identified as one of the semivolatile nuclides requiring the greatest degree of abatement prior to discharging the plant off-gas to the environment.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Review of recent developments in iodine wasteform production

Radioiodine capture and immobilization is not only important to consider during the operation of reactors (i.e., I-131), during nuclear accidents (i.e., I-131 and I-129) or nuclear fuel reprocessing (i.e., I-131 and I-129), but also during disposal of nuclear wastes (i.e., I-129). Most disposal plans for I-129-containing waste forms (including spent nuclear fuel) propose to store them in underground repositories. Here, iodine can be highly mobile and, given its radiotoxicity, needs to be carefully managed to minimize long-term environmental impacts arising from disposal. Typically, any process that has been used to capture iodine from reprocessing or in a reactor is not suitable for direct disposal, rather conversion into a wasteform for disposal is required. The objectives of these materials are to use either chemical immobilization or physical encapsulation to reduce the leaching of iodine by groundwaters. Some of the more recent ideas have been to design capture materials that better align with disposal concepts, making the industrial processing requirements easier. Research on iodine capture materials and wasteforms has been extensive. This review will act as both an update on the state of the research since the last time it was comprehensively summarized, and an evaluation of the industrial techniques required to create the proposed iodine wasteforms in terms of resulting material chemistry and applicability.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2016

A GROWTH-RATE INDICATOR FOR COMPTON-THICK ACTIVE GALACTIC NUCLEI

Due to their heavily obscured central engines, the growth rate of Compton-thick (CT) active galactic nuclei (AGNs) is difficult to measure. A statistically significant correlation between the Eddington ratio, λ {sub Edd}, and the X-ray power-law index, Γ, observed in unobscured AGNs offers an estimate of their growth rate from X-ray spectroscopy (albeit with large scatter). However, since X-rays undergo reprocessing by Compton scattering and photoelectric absorption when the line of sight to the central engine is heavily obscured, the recovery of the intrinsic Γ is challenging. Here we study a sample of local, predominantly CT megamaser AGNs, where the black hole mass, and thus Eddington luminosity, are well known. We compile results of the X-ray spectral fitting of these sources with sensitive high-energy ( E > 10 keV) NuSTAR data, where X-ray torus models, which take into account the reprocessing effects have been used to recover the intrinsic Γ values and X-ray luminosities, L {sub X}. With a simple bolometric correction to L {sub X} to calculate λ {sub Edd}, we find a statistically significant correlation between Γ and λ {sub Edd} ( p = 0.007). A linear fit to the data yields Γ = (0.41 ± 0.18)log{sub 10} λ {sub Edd} + (2.38 ± 0.20), which is statistically consistent with results for unobscured AGNs. This result implies that torus modeling successfully recovers the intrinsic AGN parameters. Since the megamasers have low-mass black holes ( M {sub BH} ≈ 10{sup 6}–10{sup 7} M {sub ⊙}) and are highly inclined, our results extend the Γ– λ {sub Edd} relationship to lower masses and argue against strong orientation effects in the corona, in support of AGN unification. Finally this result supports the use of Γ as a growth-rate indicator for accreting black holes, even for CT AGNs.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2017

PTF11kx: A Type Ia Supernova with Hydrogen Emission Persisting after 3.5 Years

The optical transient PTF11kx exhibited both the characteristic spectral features of Type Ia supernovae (SNe Ia) and the signature of ejecta interacting with circumstellar material (CSM) containing hydrogen, indicating the presence of a nondegenerate companion. In this paper, we present an optical spectrum at 1342 days after peak from Keck Observatory, in which the broad component of Hα emission persists with a similar profile as in early-time observations. We also present Spitzer IRAC detections obtained 1237 and 1818 days after peak, and an upper limit from Hubble Space Telescope ultraviolet imaging at 2133 days. We interpret our late-time observations in the context of published results—and reinterpret the early-time observations—in order to constrain the CSM's physical parameters and to compare to theoretical predictions for recurrent-nova systems. We find that the CSM's radial extent may be several times the distance between the star and the CSM's inner edge, and that the CSM column density may be two orders of magnitude lower than previous estimates. We show that the Hα luminosity decline is similar to other SNe with CSM interaction and demonstrate how our infrared photometry is evidence for newly formed, collisionally heated dust. We create a model for PTF11kx's late-time CSM interaction and find that X-ray reprocessing by photoionization and recombination cannot reproduce the observed Hα luminosity, suggesting that the X-rays are thermalized and that Hα radiates from collisional excitation. In conclusion, we discuss the implications of our results regarding the progenitor scenario and the geometric properties of the CSM for the PTF11kx system.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2016

Waste Management Approach for a Potential Future U.S. Closed Fuel Cycle - 16233

The U.S. Department of Energy, Office of Nuclear Energy has chartered an effort to develop technologies to enable safe and cost effective recycle of commercial used nuclear fuel (UNF) in the U.S. Part of this effort includes the evaluation of exiting waste management technologies for effective treatment of wastes in the context of current U.S. regulations and development of waste forms and processes with significant cost and/or performance benefits over those existing. This study summarizes the results of these ongoing efforts with a focus on the highly radioactive primary waste streams. The primary streams considered and the recommended waste forms include: •Tritium separated from either a low volume gas stream or a high volume water stream. The recommended waste form is low-water cement in high integrity containers. •Iodine-129 separated from off-gas streams in aqueous processing. There are a range of potentially suitable waste forms. As a reference case, a glass composite material (GCM) formed by the encapsulation of the silver Mordenite (AgZ) getter material in a low-temperature glass is assumed. A number of alternatives with distinct advantages are also considered including a fused silica waste form with encapsulated nano-sized AgI crystals. •Carbon-14 separated from LWR fuel treatment off-gases and immobilized as a CaCO3 in a cement waste form. •Krypton-85 separated from LWR and SFR fuel treatment off-gases and stored as a compressed gas. •An aqueous reprocessing high-level waste (HLW) raffinate waste which is immobilized by the vitrification process in one of three forms: a single phase borosilicate glass, a borosilicate based glass ceramic, or a multi-phased titanate ceramic [e.g., synthetic rock (Synroc)]. •An undissolved solids (UDS) fraction from aqueous reprocessing of LWR fuel that is either included in the borosilicate HLW glass or is immobilized in the form of a metal alloy in the case of glass ceramics or titanate ceramics. •Zirconium-based LWR fuel cladding hulls and stainless steel (SS) fuel assembly hardware that are washed and super-compacted for disposal or as an alternative Zr purification and reuse (or disposal as low-level waste, LLW) by reactive gas separations. •Electrochemical process salt HLW which is immobilized in a glass bonded Sodalite waste form known as the ceramic waste form (CWF). •Electrochemical process UDS and SS cladding hulls which are melted into an iron based alloy waste form. Mass and volume estimates for each of the recommended waste forms based on the source terms from a representative flowsheet are reported.

Fuel Cycle↗
DOE OSTI2017

De-nitration Process in Glass Melter Off-Gas Treatment System - 17092

Vitrification technology has been generally applied to stabilize High-Level Radioactive Wastes (HLW) generated from spent fuel reprocessing for safe interim storage and final geological disposal. High temperature operation of the glass melter system generates substantial amounts of NOx with various hazardous radioactive elements, due to HLW containing a lot of nitrate. Hence the off-gas treatment system (OGTS) is one of the essential systems for the removal of hazardous radioactive elements including NOx, to ensure the stable and safe operation of the glass melter system. In general, the wet scrubbing process (e.g. Sub-merged Bed Scrubber: SBS) is installed in the upstream of the OGTS for HLW vitrification in order to remove volatile ruthenium, cesium and other radioactive mists and dusts. NOx together with ruthenium, iodine and submicron particles are removed in the wet absorber equipment to below the regulated limit by law. In the Rokkasho Vitrification Laboratory (RVL) constructed recently for inactive full scale mock-up tests, we have adopted the de-nitration process using reduction catalyst in addition to together with the wet scrubbing process considering the future scenarios that are compliant to the even stricter Air Pollution Control Law. By introducing this new system, we have successfully demonstrated a good de-nitration performance in the full scale mock-up test of the glass melter system. On the other hand, by using the wet scrubbing process, nitric acid would be generated by the absorption of NOx on this recovery process and should be reworked or disposed of as secondary liquid wastes. This is one of the deciding factors for enlarging the facility and increasing construction cost. That is the reason why IHI has developed the OGTS omitting the wet scrubbing process. In this paper, we have summarized the results on the de-nitration process of the OGTS in RVL and the progress of the optimizing denitration process for future reprocessing system applications. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Ruthenium Volatilisation in Nuclear Waste Systems -Comparison of Ru(III)Cl{sub 3} and Ru(III)NO(NO{sub 3}){sub 3} - 17113

Ruthenium is a fission product possessed of two relatively long lived isotopes; {sup 103}Ru and {sup 106}Ru, both of which form part of the Highly Active (HA) waste raffinate during spent nuclear fuel reprocessing. During reprocessing, ruthenium, which may be in the form of the RuNO{sub 3}+ complex, encounters temperatures conducive to volatilisation. Due to ruthenium's high specific radioactivity it is important to understand the mechanism by which volatilisation occurs. Here we use combined cyclic voltammetry (CV) and electrochemical micro-gravimetry experiments in a study of both NO free Ru (III), in the form of a RuCl{sub 3} solution and NO complexed Ru (III), in the form of RuNO{sub 3}+ solutions. We do this in order to establish baseline behaviour of NO free Ru (III) and compare it to NO complexed Ru (III) to determine the effect of complexation on the system. We have observed discrete oxidations in both systems from solution phase Ru (III) to solid phase RuO{sub 2} and volatile RuO{sub 4}. Two key differences emerged between the two systems, the first being the mass changes associated with deposition and stripping of RuO{sub 2} are significantly smaller in the RuNO{sub 3}+ solutions than those for solutions of RuCl{sub 3}. The second difference is the difference in onset of volatile RuO{sub 4} generation, in the RuNO{sub 3}+ solution, RuO{sub 4} generation is 100 mV positive than that for the RuCl{sub 3} solution. This suggests NO complexation protects Ru (III) against oxidation and, to a certain extent, volatilisation as RuO{sub 4}. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Nuclear Liquid Wastes Calcination: The High-Level French Experience - 17184

Calcination is an efficient process for volume reduction and stabilization of nuclear liquid wastes. It is often the first step of waste containment processes such as vitrification or grouting. The CEA (French Alternative Energies and Atomic Energy Commission) and AREVA have acquired a high-level experience in the field of calcination through more than forty years of R and D and industrial operation. During this period, a broad range of liquid wastes have been studied and treated. R and D work and associated engineering studies allowed defining the design and sizing of the equipment and the process control parameters to be applied for the calcination of all kinds of solutions. The calcination process has been qualified for the following effluents amongst others: - Liquid waste from UOx fuel reprocessing; - Liquid waste from UMo fuel reprocessing (high molybdenum and phosphorus contents); - Hanford HLW; - Liquid waste from D and D operation (sodium bearing waste). This paper presents several studies led by the Joint Vitrification Laboratory (L.C.V), a common research laboratory between CEA and AREVA in charge of qualifying new processes and matrices for waste containment. These studies allowed defining the calcination rules, in particular for very hard to process solutions such as effluents with high boron, molybdenum and sodium contents. The feedback from industrial operation of the calcination process with fission products and effluents from D and D operations, highly loaded with sodium, is also presented. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Effectiveness of actinide management in fast reactors over a finite recycling period - 15123

Long-lived nuclear waste primarily consists of transuranic (TRU) isotopes. Full recycling of TRU isotopes can, in theory, lead to a reduction in repository radiotoxicity to reference levels in as little as ∼500 years provided reprocessing and fuel fabrication losses are limited. However, over a limited time frame, the radiotoxicity of the 'final' core can dominate over reprocessing losses, leading to a much lower reduction in radiotoxicity compared to that achievable at equilibrium. In this paper, the fuel cycle code ORION is used to model the recycle of light water reactor (LWR)-produced TRUs in sodium-cooled fast reactors (SFRs). With full actinide recycling, at least 6 generations of SFRs are required in a gradual phase-out of nuclear power to achieve transmutation performance approaching the theoretical equilibrium performance. Break-even SFRs are much less effective from a point of view of reducing waste radiotoxicity. Thorium (Th)-burner fuel cycles are found to result in slightly lower radiotoxicity than U-burner fuel cycles for the same number of generations. Closed break-even Th fuel cycles require ∼3 generations of operation before their waste radiotoxicity benefits result in decay to the reference level in ∼1000 years, provided that Pa is recycled. While this is a very long time frame, it is substantially better than the time frame required by other strategies. The integrated decay heat over the scenario time frame is substantially higher for break-even Th-SFRs than break-even U-SFRs as a result of much higher {sup 90}Sr production. As decay heat affects repository sizing, this weakens the argument for the Th cycle. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Progress in Defining the UK Highly Active Storage Tanks POCO Strategy - 16134

In the UK Highly Active (HA) waste is stored in stainless steel tanks prior to encapsulation in glass in one of three operating vitrification plants. Significant progress has been made in recent years towards the reduction in the volume of HA Liquor (HAL) stored in these tanks and a number of the tanks are now operating at heel levels. As a result, a strategy is being developed to carry out Post Operational Clean Out (POCO) of some tanks while reprocessing operations continue, thus enabling accelerated and more cost effective clean-up of these HA facilities. This paper, and subsequent conference presentation, will highlight some of the key aspects of this strategy, including a description of the challenges and progress made so far to help facilitate the clean-up mission without having a detrimental effect on ongoing reprocessing operations. (authors)

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

Characterization of Highly Active Cells in Preparation for Post Operational Clean Out at Sellafield - 18139

The Highly Active Liquor Evaporation and Storage (HALES) Plant at Sellafield receives highly active (HA) raffinate liquors from the Magnox and Thermal Oxide (THoRP) Reprocessing Plants. These liquors are concentrated through evaporation into Highly Active Liquor (HAL), conditioned and stored in HALES before immobilisation in glass in the Waste Vitrification Plant (WVP). Reprocessing operations at Sellafield are nearing completion and hence these liquor feeds will soon cease. Therefore, HALES is starting its planning and preparation phases for Post Operational Clean Out (POCO) and decommissioning. POCO is a transitional phase between operations and decommissioning and encompasses activities undertaken to remove residual inventory and activity from the facility. Key drivers for POCO include risk and hazard reduction and facilitating subsequent decommissioning activities (including potential surveillance and maintenance or care and maintenance regimes). Characterization is a critical enabler for development and implementation of effective POCO. In addition to the key drivers above characterization within HALES will be used to: - Define the start and end points of POCO (radiological condition, inventory sources and location, configuration, etc.); - Develop and technically underpin the wash out strategy (volumes, rates, reagent types, etc.; - Monitor the effectiveness of POCO activities (reduction in inventory/activity, points of diminishing return, confirmation of end points for POCO). There are a number of significant challenges that present to characterization of the HALES facilities. Perhaps the most significant of these is the radiological status of the facility. HALES is an HA facility meaning that the high dose rates limit not only the technology selection for characterization (i.e. those with a very high radiation tolerance) but also place constraints on the access and deployment for the selected technology. Additionally, parts of the facility are circa 60 years old which brings its own challenges such as poor quality construction drawings and limited operational knowledge. Finally, limited sampling capability and minimal data regarding historic operations has resulted in uncertainties that will need to be addressed in order to undertake POCO as efficiently and effectively as possible. The solution to this challenge requires a novel and innovative approach and an example of this is the technique which is described in this paper. This is a radiometric characterization technique which utilizes remotely deployed dosimetry to determine the activity distribution within the vessels of interest to POCO. The results to date have been extremely promising, and the ability to produce concise visual representation of these results has already begun to facilitate better informed discussions around the POCO of the facility. It is expected that in the future this technique will be applicable to other areas of Sellafield which face similar characterization challenges. This paper presents a summary of the development and implementation of the solution along with the results obtained so far. (authors)

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

Radionuclide Inventory Calculations for the Calcined Solid Storage Facility Performance Assessment - 19292

Originally named the Idaho Chemical Processing Plant when it was constructed in the early 1950's, the Idaho Nuclear Technology and Engineering Center reprocessed SNF to recover highly enriched uranium and other nuclear-related products for the U.S. DOE. Reprocessing SNF generated millions of gallons of aqueous radioactive waste. Calcination was a process developed at the INL Site to convert liquid waste into a solid, granular form from 1963 to 2000. The calcination process produced a safer product for storage while reducing the volume of stored liquid waste by an average factor of seven. Calcine is currently stored in the Calcined Solids Storage Facility (CSSF). The CSSFs are often called 'bin sets' because each CSSF (CSSFs 1 through 6) contains several stainless-steel storage bins housed in a concrete vault. The CSSF is an interim waste storage facility. In the future, calcine will be retrieved, treated (if necessary), and sent to an off-Site facility for storage or disposal. To close the CSSF, the bulk of the calcine will be removed and stored or disposed of elsewhere. The bins and vaults will then be placed in a configuration that is protective of human health and the environment. For purposes of the performance assessment, the depth of the residual calcine at closure was assumed to be 5-cm in each bin. The 5-cm of residual calcine in each bin is assumed to be mixed in grout during closure. The activities of the residual radionuclides are the inventory used for the dose assessment calculations for the CSSF performance assessment. The objective of this analysis was to calculate the radionuclide activity in the residual calcine remaining in the CSSF after closure. The purpose of this paper is to present the approach used to calculate the radionuclide inventory of the calcine currently stored in CSSF and the residual calcine in the CSSF after closure using data from the calcined waste Historical Processing Model (HPM). Limited sampling and analysis of the calcine in the CSSF bins was conducted in the past. Future similar calcine sampling and analysis efforts would be costly, generate additional mixed waste, and may not provide the desired detailed calcine composition data. However, during the later years of calcining, an alternative to calcine sampling and analysis was developed and the information was captured in the HPM. The HPM uses multiple sources to provide definitive calcine volume, mass, and composition, including: compositions and volumes of the liquid wastes that were calcined, ORIGEN2 - and ORIGEN-ARP-based radioactivity calculations, calciner and CSSF operating data, and process knowledge of INTEC fuel and waste-processing systems. HPM calcine composition was used to develop an inventory of radionuclides for use in the performance assessment evaluation for closure of CSSF. There are 43 bins that contain calcine, with six to twelve segments in each bin for a total of 337 bin segments for which the HPM calculates the average calcine composition. The HPM calculates the composition of each of the nearly 7,000 batches of material (startup bed, nonradioactive material, and radioactive waste) charged to the calciners. The calculations use the CSSF dimensions, including the CSSF number of bins, heights, outside diameters, annular air space diameter (if necessary), and volume of calcine currently in the bins. For each of the CSSFs, for a 5-cm residual, the percent of total calcine was calculated along with the residual activity for the radionuclides. Residual activity was estimated for 148 different radionuclides. (authors)

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

Panel Session 32: Panel: Bringing First-of-a-Kind Treatment Facilities On-Line: Idaho IWTU Commissioning Approach (2/5)

This panel discussed the activities of bringing a first-of-a-kind treatment facility online: The Integrated Waste Treatment Unit (IWTU). The design was designed to treat 900,000 gallons of sodium-bearing liquid waste. Construction of this first-of-a-kind treatment facility was completed in 2012. A new approach was adopted in 2016 after multiple challenges were encountered during facility commissioning. This session focused on the Integrated Waste Treatment Unit (IWTU), which was constructed to convert the remaining 900,000 gallons of liquid sodium-bearing waste at the Idaho National Laboratory Site, Idaho, USA, to a granular solid. The liquid waste was generated during historic spent nuclear fuel reprocessing campaigns at the Idaho Nuclear Technology and Engineering Center and was transferred to three underground storage tanks. The waste was not first-cycle waste but was rather the liquid rinse waste from flushing waste transfer lines between reprocessing runs. IWTU will employ a steam-reforming process to treat the liquid waste. The resulting granular solid will be transferred to stainless steel canisters and then to concrete vaults, which will be stored on site until an off-site national geologic repository is available to accept the waste. Presentation proposed: Bringing First-of-a-Kind Treatment Facilities On-Line: Idaho IWTU Commissioning Approach (Joel Case, US DOE (USA), Leo Thompson, Sterling Engineering and Consulting Group, LLC (USA), Joseph Giebel, Fluor Idaho, LLC (USA), Richard Boardman, Idaho National Laboratory (USA), Kevin O'Neill, US DOE (USA))

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

Full-Scale Retrieval and Transfer System Mockup for Pneumatic Conveyance of High-Level Calcine Waste - 19563

The Idaho National Laboratory reprocessed spent nuclear fuel from 1953 to 1994. The reprocessing used solvent extraction to separate uranium, resulting in a high-level radioactive liquid waste. A calcination process was used to convert the liquid waste to a high-level radioactive granular solid, referred to as calcine. The calcine is stored in stainless-steel bins contained within six concrete vaults called the Calcined Solids Storage Facilities (CSSFs). The Calcine Disposition Project is tasked with demonstrating the ability to retrieve the calcine from the bins of CSSF 1 and transferring it to the bins of CSSF 6. CSSF 1 was constructed to be a final calcine repository, with no forethought of retrieval. It represents the greatest technical hurdle to calcine extraction of all the CSSFs. The challenges inherent with accessing the calcine, safely extracting it, and securely transporting it are copious and complex. To overcome those challenges in a timely and objective manner, a full-scale mockup will be constructed. The mockup includes a full-size model of a concentric group of bins from CSSF 1 and a storage vessel representing CSSF 6. Mockup equipment will include an air compressor, filtration, blower/vacuum, and retrieval devices. In addition to defining process parameters, the mockup testing scope will demonstrate the amount of residual calcine that can be expected to be in the bins following extraction. This will be key to preparing for closure of CSSF 1 following removal of the calcine. Once testing is complete, the mockup will constitute a model of the calcine extraction and transport system and will be prepared to fulfill a second advantageous objective: the operators that will carry out the calcine transfer from CSSF 1 to CSSF 6 will be trained using the mockup. The control system will be identical, the major pieces of equipment will be representative, and the procedures will be analogous; hence, high-fidelity training can take place at the mockup. Once radiological operations commence at CSSF 1, the mock-up's commission will still not be completed. If any problems arise during calcine transport, the mockup will be ready to serve as a simulator to more effectively reproduce dilemmas and deduce solutions. The mock-up's mission spans the duration of the project. Its usefulness begins with helping to identify process criteria. Operations personnel will benefit greatly from its use as a training tool. And finally, the immediate availability of a simulator to reproduce problems encountered during radiological extraction and transfer will be invaluable. The knowledge gained during this process will also be of continued benefit as future work is done to remove various types of calcine from the remaining CSSFs. The methodology of safe and secure pneumatic transfer of radioactive material may be of benefit to others throughout the U.S. Department of Energy complex. (authors)

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

A Dose Assessment Model for Radioactive Waste Exposed by Gully Erosion at West Valley - 20513

The Western New York Nuclear Service Center (WNYNSC), located approximately 48 km south of Buffalo, New York, is the site of a former nuclear fuel reprocessing and radioactive waste disposal facility. Spent nuclear fuel was processed there from 1966 to 1972, leaving behind radioactive and chemical wastes in two disposal areas and a waste tank farm. Site operations also resulted in releases of radioactivity to site soils, groundwater, and to surface waters draining the site. The New York State Energy Research and Development Authority (NYSERDA) and the U.S. Department of Energy (DOE) are collaborating in a process of decision making for decommissioning those facilities remaining at the WNYNSC following the completion of Phase 1 decommissioning. Neptune and Company, Inc. (Neptune) was contracted to develop a probabilistic performance assessment (PPA) computer model to assist the agencies in this process. The PPA Model includes a contaminant transport component focusing on the movement of contaminants within and among environmental media including groundwater and surface water transport, contaminant translocation by plants and animals, diffusion, and erosion. The model also includes evaluation of potential exposure and health effects for a Resident Farmer exposure scenario, where the Resident Farmer represents a critical group, described as that group of individuals reasonably expected to receive the greatest exposure to residual radioactivity for any applicable set of circumstances. The West Valley Site is located in the glaciated Allegheny Plateau region of western New York State. The waste reprocessing and disposal areas were constructed on a relatively fat area of plateau dissected by drainages of Buttermilk Creek, including Erdman Brook, Franks Creek, and Quarry Creek. An important aspect of the contaminant transport component of the PPA model is consideration of erosive processes such as slumping of the stream slopes and the advance of gullies from these streams. These erosion processes remove material from the plateau, growing the size of the creek valleys and making them wider and deeper. Of particular interest for the impact of erosion is radioactive waste contained in the Nuclear Regulatory Commission (NRC)-Licensed Disposal Area (NDA), the New York State-Licensed Disposal Area (SDA), and residual radiological inventory in the underground storage tanks at the Waste Tank Farm (WTF). The PPA Model is organized around geographically-defined facilities which were constructed upon the plateau, including the NDA, SDA, and WTF. Ongoing stream erosion processes will potentially transfer radioactive waste and residual inventory from these facilities to the ground surface on adjacent hillslope areas where erosion has breached the facility. Hence, it is important to evaluate the consequences of potential exposures to a Resident Farmer on the hillslopes below a breached facility. Two interrelated aspects of the dose assessment model related to hillslope exposure are discussed: 1) representation of the physical processes related to transport of radionuclides from facilities onto the hillslopes, and from the hillslopes into adjoining creeks where contaminated material migrates downstream with surface water and sediment, and; 2) adaptation of the activities associated with the Resident Farmer scenario to assess potential exposures to contamination in the hillslope areas. This discussion will cover the conceptual basis of the hillslope exposure and transport models, and also implementation in the PPA computer model. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗