SSBD for Advanced Reactor Fuel Fabrication and Reprocessing.
Abstract not provided.
Cipiti, Benjamin B.↗
SEARCH · Engineering Papers
Results for “Fuel Reprocessing”
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Advanced dry head-end reprocessing of light water reactor spent nuclear fuel
A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450.degree. C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80.degree. C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Advanced dry head-end reprocessing of light water reactor spent nuclear fuel
A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450.degree. C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80.degree. C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.
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Influence of uranyl complexation on the reaction kinetics of the dodecane radical cation with used nuclear fuel extraction ligands (TBP, DEHBA, and DEHiBA)
Specialized extractant ligands – such as tri-butyl phosphate (TBP), N,N-di-(2-ethylhexyl)butyramide (DEHBA), and N,N-di-2-ethylhexylisobutryamide (DEHiBA) – have been developed for the recovery of uranium from used nuclear fuel by reprocessing solvent extraction technologies. These ligands must function in the presence of an intense multi-component radiation field, and thus it is critical that their radiolytic behaviour be thoroughly evaluated. This is especially true for their metal complexes, where there is negligible information on the influence of complexation on radiolytic reactivity, despite the prevalence of metal complexes in used nuclear fuel reprocessing solvent systems. Here we present a kinetic investigation into the effect of uranyl (UO 2 2+ ) complexation on the reaction kinetics of the dodecane radical cation (RH˙ + ) with TBP, DEHBA, and DEHiBA. Complexation had negligible effect on the reaction of RH˙ + with TBP, for which a second-order rate coefficient (k) of (1.3 ± 0.1) × 10 10 M -1 s -1 was measured. For DEHBA and DEHiBA, UO 2 2+ complexation afforded an increase in their respective rate coefficients: k(RH˙ + + [UO 2 (NO 3 ) 2 (DEHBA) 2 ]) = (2.5 ± 0.1) × 10 10 M -1 s -1 and k(RH˙ + + [UO 2 (NO 3 ) 2 (DEHiBA) 2 ]) = (1.6 ± 0.1) × 10 10 M -1 s -1 . This enhancement with complexation is indicative of an alternative RH˙ + reaction pathway, which is more readily accessible for [UO 2 (NO 3 ) 2 (DEHBA) 2 ] as it exhibited a much larger kinetic enhancement than [UO 2 (NO 3 ) 2 (DEHiBA) 2 ], 2.6× vs. 1.4×, respectively. Complementary quantum mechanical calculations suggests that the difference in reaction kinetic enhancement between TBP and DEHBA/DEHiBA is attributed to a combination of reaction pathway (electron/hole transfer vs. proton transfer) energetics and electron density distribution, wherein attendant nitrate counter anions effectively ‘shield’ TBP from RH˙ + electron transfer processes.
38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
Radiation Chemistry Research for Improved Nuclear Fuel Cycles
Nuclear energy is a key component of a global, low-carbon energy future. One of the challenges facing large-scale deployment of nuclear power is the long-term disposition of used nuclear fuel. Due to the political and technical challenges of siting deep geologic storage facilities, reprocessing of used nuclear fuel is desirable to reduce the total volume and decay heat of waste requiring long-term storage, enabling efficient usage of geologic storage repositories. However nuclear fuel reprocessing is complicated by intense multi-component radiation fields that degrade solvent systems that are used for this purpose. Resulting radiolytic degradation can result in reduced performance over time, and generation of additional waste. Thus, an understanding of the radiation-induced chemistry of these reprocessing solvent systems is crucial to for the development of new, radiation-resistant processes for efficiently treating used nuclear fuel. Controlled irradiation experiments combined with high-performance chromatography and mass spectrometry have provided comprehensive knowledge of radiolytic degradation pathways and kinetics, which is needed to predict the effect of radiation-induced chemistry on the performance of the nuclear fuel reprocessing technology. This talk will give an overview of recent progress in research on the radiation chemistry of fuel reprocessing molecules conducted by the Idaho National Laboratory Center for Radiation Chemistry Research (CR2) in conjunction with the Molecular Mass Spectrometry and Chromatography group.
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An Enhanced Safeguards Approach for Accelerator Driven Systems utilized to Close the Fuel Cycle
Further to our previous safeguards approach for Accelerator Driven Systems, which focused on estimates of fissile material production using relevant proton accelerator systems and corresponding safeguards needs for fuel storage, the subcritical reactor, and spent fuel storage material balances areas, this report is more expansive and considers utilization of ADS for either burning of transuranics or breeding of fissile materials. We find that the recycled fuels likely intended for ADS will be thermally and radioactively hot to such a degree that it is likely reprocessing and fuel fabrication will have to be co - located with the ADS reactor facility to avoid impractical hot fuel transportation issues. As such, we consider in detail the full ADS system to include material balance areas for spent fuel receiving, reprocessing, storage & cooling, fuel fabrication, subcritical reactor area, and waste storage & handling. Furthermore, aqueous - based separation methods like PUREX cannot tolerate the intense heat of the ADS fuels, so pyroprocessing will likely be required. With these considerations, we developed an Enhanced Safeguards Approach for ADS beyond the work done in our first report, and conclude that significant diagnostic development is needed , a nd provide safeguards recommendations. We have also included an appendix regarding some country programs, in particular the Chinese ADANES burner/breeder program a nd the Indian thorium - based breeder program.
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Waste Estimates for a Future Recycling Plant in the US Based Upon AREVA Operating Experience - 13206
Estimates of process and secondary wastes produced by a recycling plant built in the U.S., which is composed of a used nuclear fuel (UNF) reprocessing facility and a mixed oxide (MOX) fuel fabrication facility, are performed as part of a U.S. Department of Energy (DOE) sponsored study [1]. In this study, a set of common inputs, assumptions, and constraints were identified to allow for comparison of these wastes between different industrial teams. AREVA produced a model of a reprocessing facility, an associated fuel fabrication facility, and waste treatment facilities to develop the results for this study. These facilities were divided into a number of discrete functional areas for which inlet and outlet flow streams were clearly identified to allow for an accurate determination of the radionuclide balance throughout the facility and the waste streams. AREVA relied primarily on its decades of experience and feedback from its La Hague (reprocessing) and MELOX (MOX fuel fabrication) commercial operating facilities in France to support this assessment. However, to perform these estimates for a U.S. facility with different regulatory requirements and to take advantage of some technological advancements, such as in the potential treatment of off-gases, some deviations from this experience were necessary. A summary of AREVA's approach and results for the recycling of 800 metric tonnes of initial heavy metal (MTIHM) of LWR UNF per year into MOX fuel under the assumptions and constraints identified for this DOE study are presented. (authors)
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Performance Criteria for Capture and/or Immobilization Technologies (Revision 1)
The capture and subsequent immobilization of regulated volatile radionuclides from the off-gas streams of a used nuclear fuel (UNF) reprocessing facility has been a topic of substantial research interest for the US Department of Energy and its international counterparts. Removal of specific radionuclides from the plant effluent streams before discharge to the environment is required to meet regulations set forth by the US Environmental Protection Agency. Upon removal, the radionuclides, as well as associated sorbents that cannot be regenerated in a cost-effective manner, are destined for conversion to a waste form. Research in separation and capture methodologies has included a wide range of technology types, and studies of waste forms are correspondingly diverse. In considering the future development and implementation of both sorbents and waste forms, it is necessary to identify benchmark measures of performance to objectively evaluate each sorbent system or waste form. Sets of performance criteria and associated metrics have been developed for sorbent and waste form evaluation. These criteria address physical, radiological, and chemical characteristics, technical practicality, technical maturity, cost, and, for sorbents, system performance. The criteria and metrics appear to be robust and should be applicable despite the eventual waste classification (as either high- or low-level waste). They are flexible enough to address both aqueous reprocessing and electrochemical reprocessing of UNF. These criteria sets can serve as tools to evaluate performance at multiple stages within the development process, and in this revision (Revision 1) they have been used to assess technologies relating to krypton/xenon separations and iodine capture from off-gas streams arising from UNF reprocessing. Assessment of krypton/xenon separations using engineered forms of two zeolite minerals (silver mordenite and hydrogen mordenite in a polyacrylonitrile-based binder [AgZ-PAN/HZ-PAN]) found that the zeolite-based separation is relatively advanced in its development, but several key issues require resolution. First, desorption processes for both krypton and xenon require refinement to provide an understanding of the product purity that can be achieved. Second, adsorption rate data is needed in order to calculate the bed depth required for effective separation. Finally, it is strongly recommended that a technical review of krypton/xenon separation by AgZ-PAN/HZ-PAN be performed to synergize available data and assess the cost savings and operational benefits that may be realized from implementation of this technology. Assessment of metal organic frameworks (MOFs) for their use in the separation of krypton/xenon found that the ideal separation would be performed using a single-column system with a MOF selective for krypton over xenon. A robust research effort should work to identify a krypton-selective MOF designed to operate at temperatures of approximately 0°C or higher, which could be preferred over cryogenic krypton/xenon capture for used fuel reprocessing off-gas streams. In the case of the CaSDB-MOF (the most well-understood xenon sorbent to date), two issues are judged of high importance. First, xenon breakthrough capacity for the CaSDB-MOF in prototypical conditions should be determined. Preliminary research indicates that breakthrough may be near immediate, presenting a substantial obstacle in separative system design. Second, development of desorption methodology should be performed to determine regeneration time, energy requirements, and the product stream composition. Silver-based sorbents (AgZ and AgAero) for use in iodine capture from the dissolver off-gas were evaluated against the established criteria. These sorbents are significantly better understood for this application as a result of research efforts over the past decade. The potential implementation of AgAero at a large scale is hindered by its physical degradation by components of the dissolver off-gas stream. Less is known about the adsorption of iodine by these sorbents from other off-gas streams in the plant. Initial experimental efforts have been closely coordinated in an effort to understand organic iodine (such as would be found in the vessel off-gas) adsorption by AgZ and AgAero. Future work should expand this experimental program, and analysis of other reprocessing facility off-gas streams such as the vitrification off-gas stream should be conducted to better understand other potential applications for iodine sorbents. A review of iodine waste form development shows that this area is diverse and that multiple promising waste forms have been identified for the immobilization of radioactive iodine. Efforts related to the direct conversion of iodine sorbents (including AgZ and AgAero) should be continued because of the advantages of direct conversion in a waste management strategy and other sorbents should continue to be advanced as merited.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Assessment for advanced fuel cycle options in CANDU
The possible options for advanced fuel cycles in CANDU reactors including actinide burning options and thorium cycles were explored and are feasible options to increase the efficiency of uranium utilization and help close the fuel cycle. The actinide burning TRUMOX approach uses a mixed oxide fuel of reprocessed transuranic actinides from PWR spent fuel blended with natural uranium in the CANDU-900 reactor. This system reduced actinide content by 35% and decreased natural uranium consumption by 24% over a PWR once through cycle. The thorium cycles evaluated used two CANDU-900 units, a generator and a burner unit along with a driver fuel feedstock. The driver fuels included plutonium reprocessed from PWR, from CANDU and low enriched uranium (LEU). All three cycles were effective options and reduced natural uranium consumption over a PWR once through cycle. The LEU driven system saw the largest reduction with a 94% savings while the plutonium driven cycles achieved 75% savings for PWR and 87% for CANDU. The high neutron economy, online fuelling and flexible compact fuel make the CANDU system an ideal reactor platform for many advanced fuel cycles.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Storage and Reprocessing of Spent Nuclear Fuel
Addressing the problem of waste, especially high-level waste (HLW), is a requirement of the nuclear fuel cycle that cannot be ignored. We explore the two options employed currently, long-term storage and reprocessing.
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Development and evaluation of a silver mordenite composite sorbent for the partitioning of xenon from krypton in gas compositions
A new engineered form composite sorbent for the selective separation of xenon from krypton in simulant composition off-gas streams resulting from the reprocessing of used nuclear fuel has been developed and evaluated. A sodium mordenite powder was incorporated into a macroporous polymer binder, formed into spherical beads and successfully converted to a 9 wt.% silver form composite sorbent. The final engineered form sorbent retained the characteristic surface area indicative of sodium mordenite powder. The sorbent was evaluated for xenon adsorption potential with capacities measured as high as 30 millimoles of xenon per kilogram of sorbent achieved at ambient temperature and 460 millimoles of xenon per kilogram sorbent at 220 K. Xenon/krypton selectivity was calculated to be 22.4 with a 1020 µL/L xenon, 150 µL/L krypton in a balance of air feed gas at 220 K. Furthermore, adsorption/desorption thermal cycling effects were evaluated with results indicating sorbent performance was not significantly impacted while undergoing numerous adsorption/desorption thermal cycles.
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Efficacy of a Solution-Based Approach for Making Sodalite Waste Forms for an Oxide Reduction Salt Utilized in the Reprocessing of Used Uranium Oxide Fuel
This paper describes various approaches for making sodalite with a LiCl-Li2O oxide reduction salt used to recover uranium from used oxide fuel. The approaches include sol-gel and solution-based synthesis processes. As-made products were mixed with 5 and 10 mass% of a Na2O-B2O3-SiO2 glass binder and these, along with product without a binder, were heated using either a cold-press-and-sinter method or hot uniaxial pressing. The results demonstrate the limitation of sodalite yield due to the fast intermediate reactions between Na+ and Cl- to form halite in solution and Li2O and SiO2 to form lithium silicates (e.g., Li2SiO3 or Li2Si2O5) in the calcined and sintered pellets. The results show that pellets can be made with high sodalite fractions (~92 mass%) and low porosities using a solution-based approach and this LiCl-Li2O salt.
electrochemical↗
Iodine Pathways and Off-Gas Stream Characteristics for Aqueous Reprocessing Plants – A Literature Survey and Assessment
Used nuclear fuel is currently being reprocessed in only a few countries, notably France, England, Japan, and Russia. The need to control emissions of the gaseous radionuclides to the air during nuclear fuel reprocessing has already been reported for the entire plant. But since the gaseous radionuclides can partition to various different reprocessing off-gas streams, for example, from the head end, dissolver, vessel, cell, and melter, an understanding of each of these streams is critical. These off-gas streams have different flow rates and compositions and could have different gaseous radionuclide control requirements, depending on how the gaseous radionuclides partition. This report reviews the available literature to summarize specific engineering data on the flow rates, forms of the volatile radionuclides in off-gas streams, distributions of these radionuclides in these streams, and temperatures of these streams. This document contains an extensive bibliography of the information contained in the open literature.
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Distribution Development for Residual Inventory at the New York West Valley Site - 20400
The New York State Energy Research and Development Authority (NYSERDA) is the owner of the Western New York Nuclear Service Center (WNYNSC), a 1,351 ha site located approximately 48 km south of Buffalo, New York. In 1962, Nuclear Fuel Services, Inc. (NFS) entered into Agreements with the Atomic Energy Commission and New York State to construct the first commercial reprocessing plant of nuclear fuel in the United States. NFS, a private company, built and operated the spent fuel reprocessing plant and waste disposal facilities, processing 640 Mg of spent nuclear fuel from 1966 to 1972 under an Atomic Energy Commission license. Nuclear fuel reprocessing operations ended in 1972 and never reopened, leaving behind radioactive and chemical wastes. Operations led to contamination in a number of facilities and locations. Some of that contamination has migrated from waste disposal zones to other layers, formations, and features on and off the WNYNSC. Phase I decommissioning activities are ongoing and involve the removal of a number of areas and structures that have been associated with contamination. The purpose of this work is to outline the approach for characterizing contamination not associated with disposed wastes, contaminated structures, or specific releases. In this work, the term, residual radiological activity, is used to describe environmental contamination that exists subsequent to the completion of Phase I decommissioning activities, that is not associated with disposed wastes, contaminated structures, or specific releases. Contamination from the Site was quantified relative to data that characterize the concentrations of radionuclides that exist in background. Background concentrations are those present in the area but having no influence from Site related activities. The existence of residual radiological activity that is elevated relative to background has the potential to contribute to future risks to human health and the environment. As a consequence, the residual inventory information is used to inform the West Valley Probabilistic Performance Assessment (PPA) model to characterize potential future risks to human health and the environment. The centralized West Valley Data Management System (DMS) was the source of information for the data assembled in this analysis. The DMS is a fairly large compilation consisting of thousands of records from investigation studies, with sample dates ranging from 1990 to present. Samples from monitoring wells, boreholes, geoprobe studies, surface water, surface soils, storm water outfalls, ventilation stack filters, plant and animal tissues, and more are included in the DMS. Results are typically reported in units of activity per unit volume. For the purpose of the analyses presented here, all results were converted into consistent units of pCi per unit volume. Since 1990, data have been collected from various locations across the WNYNSC at different times with varying frequency over the course of several decades. As a consequence, a number of potential issues can arise with respect to the assembly of a dataset that is deemed adequate for the characterization of residual radiological activity. These issues were assessed and resolved to the extent possible through careful consideration of the properties of the distributions. The intent was to use data which characterize the current state of the Site. Radionuclides can be designated to one of several groups depending on their origin. In this work the groups considered were 1) Naturally Occurring Radioactive Material (NORM), 2) fallout, and 3) Other (including power plant, medical research, etc). This grouping is a useful construct with respect to the interpretation of fixed laboratory results. For example, NORM radionuclides that exist within a decay chain should have approximately equivalent distributions of concentrations if they are representative of background conditions. Insights such as these can be used as a check to identify sample results that need to be further investigated or omitted due to issues associated with reported results from fixed laboratory analyses. This type of analysis provided a foundation for the assessment of the adequacy of sample results for use in subsequent components of an assessment. The general process for the assessment of residual radiological contamination at the Site consists of a sequence of several steps. First, for each analyte, several statistical tests were performed to assess the weight of evidence against the null hypothesis that the mean of the distribution of concentrations was equal to zero. If the mean of the distribution of concentrations for a given radionuclide was not found to be greater than zero, then it was removed from consideration as a component of the residual radiological contamination. If there was significant evidence to reject the hypothesis of the mean being equal to zero, the second step was to compare the distribution of the data from the Site to that of the corresponding background. A suite of tests was used to compare the distributions of the site and background data. The results of these tests were collectively used to determine if site data are elevated relative to background. The third step was to develop distributions using a Bayesian framework to characterize the distribution of mean of the increment present above background for each of the radionuclides. The Bayesian model implemented allowed for the comparison of site-specific records to background concentrations to better approximate contamination attributed to the Site. A final screening step was employed for radionuclides that exceed background. This screening step compared 95% upper confidence limits (UCLs) from the increment distribution developed in the previous step to the risk screening levels. This approach yields a list of analytes that were determined to be elevated relative to background.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Quantifying the impact of climate change on erosion - 20397
The New York State Energy Research and Development Authority (NYSERDA) is the owner of the Western New York Nuclear Service Center (WNYNSC), a 1,351-ha (3,338-ac) site located approximately 48 km (30 mi) south of Buffalo, New York. In 1962, Nuclear Fuel Services, Inc. (NFS) entered into Agreements with the Atomic Energy Commission and New York State to construct the first commercial reprocessing plant of nuclear fuel in the United States. NFS, a private company, built and operated the spent fuel reprocessing plant and waste disposal facilities, processing 640 Mg (metric tons, or 705 short tons) of spent nuclear fuel from 1966 to 1972 under an Atomic Energy Commission license. Nuclear fuel reprocessing operations ended in 1972 and never reopened, leaving behind radioactive and chemical wastes. Erosion can play an important role in the fate and transport of waste at sites where disposal of long-lived waste is anticipated. The statistical characterization of key processes related to erosion is essential to the understanding of site stability through time. One of the key processes governing erosion is extreme precipitation and it is critical that trends in the distribution of extreme precipitation events through time be represented. The evidence of climate change is increasingly well documented and projected impacts on extreme precipitation events should be incorporated in performance assessment studies when relevant. Not evaluating future climate states in a performance assessment is contradictory to good modeling practice. Specifically, excluding climate change limits development of modeling information that could aid in effective decision making. The current climate literature provides both observational evidence and climate model projections of climate trends and/or climate change in the late 20. and early 21. centuries for North America and the northeast United States. In this work, this information was used to assess the impacts of potential changes in climate on erosion processes. The goal was to understand how projections of future climate relate to the performance of the WNYNSC through time. In the first stage of this work multi-temporal historical aerial images were analyzed in conjunction with orthophotography and Lidar data to develop probability distributions for variables representing important erosion processes. The information from these analyses was then used in conjunction with simulated data from the West Valley Erosion Working Group (EWG). Analysis of EWG simulations provides estimates for the change in erosion rates through time that is driven by changes in climate. The time-varying rates of erosion change were applied to the historical aerial imagery data in order to inform time-varying rates of erosion that are driven by changes in climate. Ultimately, this process resulted in the identification of locations for features like gully heads using both the Lidar dataset and projection of the estimated location from the historical aerial photo under consideration back to the Lidar dataset. The distance between the estimate of the location from the historical image and that of the Lidar dataset was the estimated distance the feature has moved. This distance was then divided by the number of years between the Lidar dataset and the year of the historical aerial image of interest to get a rate of movement through time. This was done for several dozen points on each historical aerial image. The analyses from the EWG were used to characterize the relative impact of climate change on the erosion rates. This relative impact was quantified by comparing the LEM based estimates of erosion that were derived using historical climate data with LEM based estimates of erosion that were derived using projections of future climate. The relative increase in the erosion rates was then applied to the historical estimates of erosion derived from the analysis of the historical aerial imagery. This approach was used since the LEM-based estimated of the historical erosion rates from the past century have a bias towards underestimating erosion. This underestimation is hypothesized to be a consequence of an inability of the LEMs to account for the impacts of land-use change (i.e. deforestation) which had been shown to significantly increase erosion in similar Northern hardwood forest ecosystems. In summary, gully head retreat rate and gully widening rate were characterized using statistical probability distributions from the historical aerial image analysis. Simulated data from the EWG was used to estimate the climatically-driven changes in erosion rates through time. The changes through time from the EWG were applied to the gully head retreat rate and gully widening rate characterized using the historical aerial image analysis. This information can be used to inform PPA models to project future risks from a given site. (authors)
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A Preliminary Probabilistic Performance Assessment Model for Radiological and Chemical Contamination at the West Valley Site, New York - 19295
The New York State Energy Research and Development Authority (NYSERDA) is the owner of the Western New York Nuclear Service Center (WNYNSC), a 1,351-ha site located approximately 48 km south of Buffalo, New York. In 1962, Nuclear Fuel Services, Inc. (NFS) entered into agreements with the Atomic Energy Commission and New York State to construct the first commercial reprocessing plant of nuclear fuel in the United States at the WNYNSC. NFS built and operated the spent fuel reprocessing plant and waste disposal facilities, processing 640 Mg (640 metric tons) of spent nuclear fuel from 1966 to 1972 under an Atomic Energy Commission license. Nuclear fuel reprocessing operations halted in 1972 and never restarted, leaving behind radioactive and chemical wastes. The reprocessing wastes include high-level waste (HLW), so U.S. Department of Energy (DOE) was compelled to complete certain waste management activities under the West Valley Demonstration Project (WVDP) Act of 1980, including decommissioning of WVDP facilities. As collaborating agencies, NYSERDA and the DOE are tasked with making decisions about decommissioning and risk reduction for the West Valley Site. Neptune and Company, Inc. (Neptune) was contracted to develop a probabilistic performance assessment (PPA) model to assist the agencies in their decision making process for decommissioning the WVDP and WNYNSC. One important tool that is needed in order to inform the decision-making process is a science-based model of the contamination at the West Valley Site and its potential implications for human health and the environment. The West Valley PPA Model (the Model), developed using the GoldSim modeling software, is a tool intended to provide support for decision making with uncertainty, in a manner that is transparent, defensible, and robust. The PPA Model includes contaminant transport and health effects components, and is organized around geographically-organized contaminated facilities. These include the waste disposal areas licensed by the U.S. Nuclear Regulatory Commission and the State of New York, a waste tank farm for storage of HLW resulting from reprocessing operations, and several areas contaminated with radioactive and chemical constituents. The Model evaluates contaminant transport from these sources to points of exposure across the site and into receiving surface waters and sediments. Hypothetical people and wildlife are exposed to contamination at these locations, and the effects of these exposures are evaluated. Contaminant transport processes to be evaluated in the Model include groundwater and surface water transport, contaminant translocation by plants and animals, diffusion in subsurface air and water, and erosion of the Site. The evaluation of exposures to people in this preliminary model is limited to a resident farmer scenario, and ecological assessment is performed at the level of a screening analysis. The results of these preliminary evaluations inform future model developments. Model results are subjected to sensitivity analysis in order to determine those pathways and parameters that are most significant in influencing the results. This information allows analysts and decision makers to focus on those aspects of Site behavior and processes that contribute most to overall uncertainty. With this information, the Supplemental Environmental Impact Statement (SEIS) can inform defensible decisions regarding remediation of the Site. (authors)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Updates to a Preliminary Probabilistic Performance Assessment Model for Radiological and Chemical Contamination at the West Valley Site, New York - 20420
The New York State Energy Research and Development Authority (NYSERDA) is the owner of the Western New York Nuclear Service Center (WNYNSC), a 1,351-ha site located approximately 48 km south of Buffalo, New York. In 1962, Nuclear Fuel Services, Inc. (NFS) entered into agreements with the Atomic Energy Commission and New York State to construct the first commercial reprocessing plant of nuclear fuel in the United States at the WNYNSC. NFS built and operated the spent fuel reprocessing plant and waste disposal facilities, processing 640 Mg (640 metric tons) of spent nuclear fuel from 1966 to 1972 under an Atomic Energy Commission license. Nuclear fuel reprocessing operations halted in 1972 and never restarted, leaving behind radioactive and chemical wastes. The U.S. Department of Energy (DOE) was required to complete certain waste management activities under the West Valley Demonstration Project (WVDP) Act of 1980 including decommissioning of WVDP facilities. As collaborating agencies, NYSERDA and the DOE are tasked with making decisions about decommissioning and risk reduction for the West Valley Site. Neptune and Company, Inc. (Neptune) was contracted to develop a probabilistic performance assessment (PPA) model to assist the agencies in their decision making process for decommissioning the WVDP and WNYNSC. One important tool that is needed in order to inform the decision-making process is a science-based model of the West Valley Site that evaluates potential future consequences for human health and the environment. This forms the core of the spatial domain of the West Valley PPA Model. The PPA Model, developed using the GoldSim system modeling software, is a tool intended to provide support for decision making that evaluates uncertainty, in a manner that is transparent, defensible, and robust. The PPA Model includes contaminant transport and health effects components, and is organized around geographically-grouped contaminated facilities. These include the waste disposal areas licensed by the U.S. Nuclear Regulatory Commission and the State of New York, a waste tank farm for storage of high level radioactive waste resulting from reprocessing operations, and several areas contaminated with radioactive and chemical constituents. The PPA Model evaluates contaminant transport from these sources to points of exposure across the site and into receiving surface waters and sediments. Hypothetical people and wildlife could be exposed to contamination at these locations, and the effects of these exposures are evaluated. Contaminant transport processes to be evaluated in the PPA Model include groundwater and surface water transport, contaminant translocation by plants and animals, diffusion, and erosion. The evaluation of exposures to people in this preliminary model is limited to a resident farmer scenario, and ecological assessment is performed at the level of a screening analysis. The results of these preliminary evaluations inform future model developments. PPA Model results are subjected to sensitivity analysis in order to determine those pathways and parameters that are most significant in influencing the results. This information allows analysts and decision makers to focus on those aspects of Site behavior and processes. With this information, the decision makers can drive informed, defensible decisions regarding decommissioning of the Site. This paper includes an update of the information presented at WM2019 [1]. (authors)
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Iodobutane Deep Bed Adsorption Test Report
The Department of Energy (DOE) Nuclear Technology Research and Development (NTRD) Program Material Recovery and Waste Form Development (MRWFD) Campaign Off-gas Sigma Team has supported research and development on iodine control and iodine waste forms for the past several years. As a part of this on-going research, two deep-bed iodobutane adsorption tests have been performed with higher NO x levels. High (>1,000) iodine decontamination factors (DFs, the ratios of the inlet iodine concentrations divided by the outlet iodine concentrations, when the gas flowrate is constant) were achieved with both the silver aerogel and silver zeolite sorbents. The iodine adsorption mass transfer zone (the depth of the sorbent bed in which adsorption is occurring at any specific time) for these tests was estimated at 4 inches for both sorbents under the test conditions. Silver utilization for silver aerogel sorbent reached 41% (up to 12 g adsorbed iodine per 100 g silver aerogel sorbent), while silver utilization reached 43% (up to 6.8 g adsorbed iodine per 100 g silver zeolite sorbent). These values are generally consistent with results for prior methyl iodide adsorption tests. The results also indicate that, during iodine adsorption process, the organic iodide molecule decomposes so that the iodine can be adsorbed. Even when sorbent breakthrough is reached, the organic iodide molecule continues to decompose; and the unadsorbed iodine that passes through bed is no longer in the form of the original organic iodide; but in the form of iodine compounds that are soluble in NaOH solutions, such as HI or I 2 . This research and development has included iodine adsorption tests using a laboratory-scale test system containing segmented fixed beds of iodine adsorbents in series. Non-radioactive synthetic gas mixtures were blended from air, nitrogen, NO, NO 2 , water, and the target iodine compounds to represent dissolver off-gas (DOG) streams from used nuclear fuel aqueous reprocessing.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗