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At least 649 records · Page 36
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

A Decade of Innovative Approaches to Treating and Dispositioning Radioactive Waste at the Savannah River Site - 20620

This year marks the 70. anniversary of the U.S. Department of Energy's (DOE) Savannah River Site (SRS), located near Aiken, South Carolina. SRS is a key industrial complex responsible for environmental stewardship, environmental cleanup, waste management, and disposition of nuclear materials. SRS also continues to have a role in critical defense-related activities and the reprocessing of used reactor fuel. The SRS encompasses 803 square kilometers (310 square miles) in parts of Aiken, Barnwell, and Allendale counties. The Liquid Waste Mission starts with the safe receipt and storage of radioactive liquid waste in the waste tanks, which principally includes Cold War legacy waste, but also waste from support of National Aeronautics Space Administration missions, medical isotope production, and research activities. This 133,000 cubic meters (m{sup 3}) (35 million gallons [Mgal]) of high-level radioactive waste (HLW) is currently held in 43 large underground waste tanks. The capacity of each of these tanks range from 2,800 m{sup 3} to 5,000 m{sup 3} (0.75 to 1.33 Mgal) and were placed in operation between 1954 and 1986. In July 2009, the DOE awarded a stand-alone contract to execute the Liquid Waste Mission at SRS. The Liquid Waste Contractor selected was Savannah River Remediation LLC1 (SRR), which is responsible for receipt, storage, retrieval and treatment of all HLW, disposal of the decontaminated low-activity waste fraction, as well as operationally closing cleaned HLW tanks. Since July 2009, SRR has successfully grouted and operationally closed six large underground tanks, poured 1,476 canisters of vitrified HLW (a stable glass waste form), treated approximately 34,200 m{sup 3} (9.0 Mgal) of HLW, and dispositioned over 41,200 m{sup 3} (10.9 Mgal) of decontaminated salt solution as low-level waste (LLW) into the Saltstone Disposal Units (SDUs) being constructed by SRR on site. The innovative work performed by SRR in the LLW portion of its mission has earned it the Richard S. Hodes Award. The work conducted by SRR is the same work championed by the late Richard S. Hodes, a respected physician, statesmen, and Chairman of the Southeast Compact Commission for Low-Level Radioactive Waste Management. (authors)

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DOE OSTI2020

Evaluation of Material Balance Approaches for Hanford Direct-Feed Low Activity Waste Processing - 20022

The Hanford Site has accumulated millions of gallons of tank waste from reprocessing spent fuel to recover plutonium, uranium, cesium, and strontium. The supernatant from the accumulated tank waste will be treated using a Direct Feed Low-Activity Waste approach. The supernatant will be treated to remove solids and cesium in the Tank-Side Cesium Removal process in the Hanford tank farm, then vitrified in a semi-batch process in the Hanford Waste Treatment and Immobilization Plant (WTP). Currently, each batch of feed is sampled at three locations prior to being fed to the melter: the feed qualification tank in the Hanford tank farm as well as the concentrate receipt vessel (CRV) and melter feed preparation vessel (MFPV) in the WTP. The feed qualification sample is taken from a large batch of accumulated feed, only two to three samples are expected each year. Approximately 275 samples from the CRVs and 1100 samples from the MFPV are expected each year. An evaluation was performed to determine if a material balance based on the feed qualification sample could replace most of the sampling in the CRVs and MFPVs. The evaluation consisted of three elements: (1) determination of the practicality of using a material balance to estimate the stream composition of the CRV and MFPV contents, (2) evaluation of whether the material balance could be automated using the existing process control system, and (3) determination of the uncertainty in glass composition using the material balance approach. It is assumed that periodic sampling at the CRV and MFPV would be performed periodically to re-baseline the material balance, evaluations are in progress to determine the frequency of this periodic sampling. Process sample locations downstream of the melter were reviewed as well, but the partitioning of semi-volatile species in the melter was determined to preclude extending the material balance approach past the melter. It was determined that replacement of the CRV sample location was feasible and did not increase process uncertainty or significantly impact waste loading. Replacement of the MFPV sample was also determined to be feasible, but that measurement of the glass former chemical addition may be needed prior to addition of these chemicals to the MFPV. This measurement could be performed by an in situ laser-induced breakdown spectroscopy (LIBS) system. Limited tests were performed to evaluate LIBS for direct measurements of the low-activity waste melter feed. The use of a material balance would eliminate over 1200 samples each year if only 10% of the CRV and MFPV batches are sampled and could likely allow the WTP laboratory to operate on days only versus 24/7 operation. (authors)

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DOE OSTI2020

Progress towards the POCO of UK Highly Active Storage Tanks - 20112

In the UK Highly Active (HA) waste is stored in stainless steel Highly Active Storage Tanks (HASTs) prior to encapsulation in glass in one of three operating vitrification plants. Significant progress has been made towards the reduction in the volume of HA Liquor (HAL) stored in these HASTs and several tanks currently operate at heel levels. In recent years a strategy has been developed towards the Post Operational Clean Out (POCO) of the tanks and significant progress has been made in confirming this strategy since it was last presented at WMS [1]. For a variety of reasons changes to associated operational assumptions have also been made. This paper will highlight recent progress and changes associated with the strategy, including a description of some of the challenges that have been overcome and associated technical work undertaken. It will also detail current plans to POCO the HASTs, as quickly as possible, while minimising the risk and continuing to support other ongoing reprocessing and clean-up programmes as part of the overall UK clean-up mission. (authors)

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DOE OSTI2020

Estimation of Sorption Behavior of Europium to Calcium Silicate Hydrate by Thermal Analysis - 20172

High-level radioactive waste generated from reprocessing processes will be disposed in a geological disposal that is deep underground. Calcium silicate hydrate is formed as a secondary mineral in the repository via the reaction of calcium ions leaching from the cement and the silicic acid dissolved from the host rocks. In this study, the sorption behavior of europium, which is also known as a chemical analog of americium, on calcium silicate hydrate was examined. The solid phase was analyzed via thermogravimetry- differential thermal analysis, differential scanning calorimetry, Raman spectroscopy, fluorescence spectroscopy, and fluorescence lifetime analysis. To understand the behavior of calcium silicate hydrate under conditions saturated with groundwater, the study samples were analyzed using the above methods, except for deferential scanning calorimetry, without a drying process. The results of the differential scanning calorimetry analyses showed that the temperatures for the endothermic peaks of calcium silicate hydrate for the lower Ca/Si ratios were larger than those for the higher Ca/Si ratios for coexistent conditions of europium ions. Hence, the relationships of the temperatures of the endothermic peaks and the Ca/Si ratios were reversed by the addition of europium ions. These study results suggest that calcium silicate hydrate can immobilize europium ions, thus preventing the production of colloidal forms. (authors)

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DOE OSTI2020

The TRANSCEND Consortium - In-situ Identification of Surface Corrosion Products on Spent Nuclear Fuels - 20276

The management of spent nuclear fuel is a major ongoing concern for the UK owing to the cessation of reprocessing operations at Sellafield and the large, complex inventory arising from Magnox, AGR, PWR and prototype reactors. Retrieval and relocation operations for legacy fuels are imminent and therefore, any models that enhance our understanding of fuel evolution will help mitigate the risks associated with fuel storage and disposal. The TRANSCEND Consortium on nuclear waste management comprises four work packages, within this current paper we provide a summary overview of progress to date and illustrative results from Theme 3: Spent Nuclear Fuels. (authors)

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DOE OSTI2020

Aging of Concrete for the Evaluation of Repair Materials to Protect the Walls of the HCAEX Tunnel at Savannah River - 20301

H-Canyon, located in the Savannah River site, is a unique facility for chemical reprocessing of plutonium, highly-enriched uranium, and other radioactive materials. The exhaust gases of the H-Canyon are sent through the H-Canyon Exhaust (HCAEX) tunnel for contamination removal. Robotic inspections of the tunnel revealed significant degradation of the reinforced concrete structure that was associated with acid attack, and could compromise the structural stability of the tunnel. Thus, the identification and evaluation of potential repair materials that could be applied on the degraded walls to mitigate and prevent further degradation is of significant interest to the Department of Energy and the Savannah River representatives. This research effort has been divided into two phases: 1) Development and evaluation of aged concrete under accelerated aging conditions (which is the focus of this paper) and 2) Evaluation of potential repair materials applied on aged and non-aged concrete under simulated aggressive conditions. In order to develop and evaluate concrete samples exposed to accelerated aging conditions in simulated aggressive environments, a literature review of the HCAEX tunnel was conducted that included 1) characterization and extent of the concrete damage, 2) environmental conditions inside the tunnel and 3) primary deterioration mechanisms. In addition, potential coatings and/or repair materials for degraded concrete surfaces exposed to aggressive environments (primarily acidic) were selected from the literature review and the most common testing and measurements for evaluating acid attack phenomena, erosion, etc. were reviewed. From the literature review findings, a preliminary bench-scale test plan for the concrete aging was developed, including accelerated aging tests with aggressive conditions (acid fumes, humidity, etc.). Concrete samples were exposed to the aging accelerated conditions (e.g. immersion in acid solutions) and visual inspection, mass loss and pH change (acid solution) were recorded over time. Correlations between the visual inspection, mass loss and pH changes results and the aging time or the aging conditions were developed. Specimens submitted to the highest acid concentration showed the fastest and most intense degradation. The type of coarse aggregate (limestone) used for the concrete seemed to be the cause of the fastest aging observed, compared to the cement paste. The research findings created the foundation for the ongoing investigation, in which new concrete samples with a mix design similar to the HCAEX tunnel will be tested and will serve as the substrate for testing the selected coatings and /or repair materials. In this paper, the literature review and preliminary results from the aging tests are provided. (authors)

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DOE OSTI2020

Treatment of Problematic Reactive Metal Wastes Using the GeoMelt{sup R} In-Container Vitrification (ICV{sup TM}) Process - 20326

Decommissioning of sodium-cooled reactors and fast reactor technologies has generated a number of reactive metal waste configurations that are problematic to treat and typically lack cost effective treatment methods and disposition options. As a result, Veolia Nuclear Solutions, under contract with Idaho National Laboratory (owned by the U.S. Department of Energy and managed and operated by Battelle Energy Alliance, LLC) demonstrated its GeoMelt{sup R} In-Container Vitrification (ICV){sup TM} technology to safely convert sodium metal to a non-reactive vitrified oxide form. The demonstration project, supported by glass formulation and crucible testing, consisted of a series of ICV{sup TM} melts that processed elemental sodium into stable non-reactive glass. INL is currently implementing GeoMelt{sup R} technology as a means to safely and reliably convert radioactive reactive metal residues that contaminate sodium cooled reactor components into waste forms that comply with existing disposition pathways. Reactive metal wastes require treatment in order to remove the Resource Conservation and Recovery Act (RCRA) reactivity and ignitability characteristics to comply with land disposal restrictions. GeoMelt{sup R}, which is an alternative to other potential treatment approaches, provides a robust approach that chemically converts the reactive metals to an inert oxide while also immobilizing radionuclides in a vitrified waste form with durability equal to or better than vitrified nuclear fuel reprocessing wastes (very robust and inert waste forms). Most other treatment approaches generate hydrogen gas which is problematic. In 2016, Veolia Nuclear Solutions first demonstrated the effectiveness of the GeoMelt{sup R} ICV{sup TM} process in deactivating reactive sodium metal. Crucible, bench-scale, and engineering-scale demonstrations were conducted on several surrogate waste configurations with various ratios of sodium metal and glass formers. Each ratio and configuration demonstrated complete deactivation of the surrogate sodium metal. Follow-on work in 2017 demonstrated the deactivation of reactive sodium by GeoMelt{sup R} ICV{sup TM} at a higher waste loading relative to previously demonstrated work performed in 2016; the higher waste loading optimized glass chemistry while enhancing the economical full-scale treatment of reactive metals. Additionally, follow-on demonstration testing in 2018 and 2019 focused on more complex shapes and other reactive-metals (mocked up Experimental Breeder Reactor II [EBR-II] subassembly, sodium filled heat exchanger, and a can containing sodium potassium alloy) which were all performed at engineering scale. Veolia Nuclear Solutions designed, installed, and commissioned in September 2018, at Perma-Fix Northwest in Richland Washington, a 10-metric ton full-scale GeoMelt unit (GeoMelt{sup R} Richland) for the treatment of reactive metal wastes. As of September 2019, over 900 55-gallon drums containing a total of around 3,500 lb of sodium with low levels of radioactivity have been treated at GeoMelt{sup R} Richland, with resulting glass monoliths disposed at the Nevada National Security Site (NNSS). A full-scale radiological demonstration melt on an actual EBR-II subassembly has also been performed using the full-scale melter in 2019. The GeoMelt{sup R} technology is a proven radioactive waste treatment technology capable of immobilizing radioactive wastes, including bulk rubble such as drums and other steel vessels usually without pretreatment. Utilizing the GeoMelt{sup R} technology to treat reactive metals eliminates pretreatment steps resulting from having to separate the reactive metal from steel containers or jackets as GeoMelt{sup R} can easily operate at temperatures sufficient to melt the steel and expose the reactive metal for treatment. Eliminating handling steps of reactive metals is a significant safety advantage since reactive metals are pyrophoric. The results generated as a part of the 2018-2019 demonstration program are presented in the paper. (authors)

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

Near-Term Deployment Viability of Liquid-Fuel Molten Salt Reactors

Molten salt reactors (MSRs) with liquid fuel offer many benefits such as continuous fuel reprocessing, which provides unprecedented flexibility in fuel cycle selection, and excellent performance both in thermal efficiency and intrinsic safety. However, challenges such as the need for proliferation-resistant design and a high-temperature, corrosive environment for structural materials have hindered their development. New MSR concepts, such as the ThorCon Do-able Molten Salt Reactor, aim to simplify design and operation in order to cut costs and enable deployment in a relatively short period of time. The proposed methods for reducing the impact of longstanding issues on MSR deployment have incentivized further technical assessment of the reactor design. This study selectively examined the (1) thermal hydraulics, (2) neutronics, (3) primary safety systems, and (4) economics of a representative liquid-fuel MSR to better understand the technical and economic viability for near term deployment of this technology. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

A Scintillation-Fiber Based Cosmic-ray Muon Tomography System for Imaging Dry Storage Cask

Because of the growth of the nuclear power industry in the United States and the policy decision to ban reprocessing of commercial spent nuclear fuel, the spent fuel inventory at commercial reactor sites has been increasing. Spent nuclear fuel presents extra complexity of security and proliferation issues. These fuel assemblies are practically inaccessible for inspection purposes, as reopening a DSC would require special facilities and would be tremendously expensive. Cosmic-ray muon Tomography has been studied in the past for various applications and was demonstrated to be a good non-destructive assay technique for high-Z materials, such as spent nuclear fuel assemblies in dry storage casks (DSC). Compared to conventional imaging techniques, muon imaging has some advantages due to its negligible health risk and high penetrating power. In this paper, we report our simulation study on imaging a DSC using muon Tomography technique. The Point of Closest Approach (POCA) and the Maximum-Likelihood Expectation- Maximization (ML-EM) algorithms are implemented and evaluated. A scaled-down prototype muon tomography system is proposed and studied. (authors)

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

Product Consistency Testing of Hot Isostatically Pressed Iodine-containing Waste Forms

Silver-exchanged mordenite (AgZ) is currently under evaluation for the capture of iodine from the off-gas streams that are expected to arise during the reprocessing of used nuclear fuel. A final waste form for AgZ loaded with radioactive iodine has not yet been identified. The applicability of hot isostatic pressing (HIPing) to convert iodine-loaded, reduced silver-exchanged mordenite (I-Ag{sup 0}Z) into a suitable waste form is currently being evaluated at Oak Ridge National Laboratory (ORNL). Because production of such a waste form requires little preprocessing of the I-Ag{sup 0}Z, HIPing could prove very attractive from both safety and economic perspectives. It is desirable to produce a waste form with a distinct iodine-containing mineral phase as the major constituent. Such a mineral phase is thought to increase long-term iodine retention and overall stability. Furthermore, the transformation of the material into a known iodine-containing mineral phase would limit the additional experimental testing and modeling required to determine the long-term stability of the pressed pellet, as much of that information has already been learned for several common iodine-containing minerals. For these reasons, the current work has investigated the possibility of converting I-Ag{sup 0}Z into an iodine-containing sodalite, similar to the efforts documented by Sheppard et al. in which they transformed other silver zeolite structures into sodalite through HIPing. Building on our initial work we are now producing hot isostatically pressed waste forms using mixtures of pure sodium mordenite, AgZ, and silver iodide. The composition of these samples is well known, in contrast with the previously used commercially available, binder-containing AgZ, which has a composition that is proprietary and uncharacterized. It is expected that by controlling the initial sample composition, we might be able to control the mineral phases present in the processed waste form. Transformation of the component material into a well-characterized iodine-containing mineral phase would be desirable; however, this is not an absolute requirement for success. (authors)

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DOE OSTI2016

Impact of Using Cross Sections Versus Recipes in Transition Fuel Cycle Analysis using ORION

ORION, a nuclear fuel cycle simulator developed at the UK National Nuclear Laboratory (NNL), is designed to simulate the operations of general nuclear fuel cycles, including once-through, limited recycle, and continuous recycle fuel cycle options. ORION can simulate the full range of nuclear-related facilities e.g. interim and long-term storage locations, fabrication and enrichment plants, reprocessing facilities, and reactors. It can track over 2,500 nuclides and can model decay and in-reactor irradiation. Based on ORION's ability to track the full isotopic mass flow between fuel cycle facilities (with or without decay), it is possible to model all of the necessary nuclide production and destruction routes. This is essential when calculating accurate spent nuclear fuel (SNF) compositions and associated metrics (e.g., decay heat and radioactivity). When modeling the transition from the current once through US fuel cycle to an alternative fuel cycle, it is important to accurately model the US pressurized water reactor (PWR) discharged SNF composition while minimizing the computational time. This paper discusses two rapid methods used to model the US PWR discharged SNF-fixed recipes versus cross sections-and it discusses the differences between the results from these two methods in relation to fuel cycle transition analysis. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Material Balance Area Design for the Transuranic Fuel Cycle Employing High Temperature Gas Cooled Reactors

A major impetus has been to ensure safety in nuclear power plant systems from the beginning of its development to the current period of nuclear renaissance. As one of the next generation reactors with a high level of nuclear safety, high temperature gas cooled reactor (HTGR) has attracted large attention because of the possibility to shut down the reactor safely and remove decay heat without the use of active cooling. A deep-burn concept of fuel utilization is proposed and studied for HTGRs to alleviate plutonium proliferation concern. The idea is to utilize the plutonium extracted by reprocessing the light water reactor (LWR) spent fuel and discharge it only at higher fuel burn-ups of up to 90 atom %. In addition HTGRs have been studied to utilize transuranic (TRU) fuel. A large amount of special nuclear material (SNM) will be used in the HTGR advanced fuel cycle. The inherent proliferation resistance and material attractiveness of the fresh fuel and discharged spent fuel have been discussed against the diversion from peaceful to military uses. On the other hand safeguards of SNM is also important to prevent nuclear proliferation under the Non-Proliferation Treaty between the state and International Atomic Energy Agency (IAEA). Facility operators are responsible to design material balance area (MBA) and provide enough information on material balance of SNMs for the safeguards conclusion. The material unaccounted for (MUF) and measurement uncertainty in MUF (σ{sub MUF}) represent accuracy of material balance. The MBAs have to be designed and established to achieve enough confidence level on MUF and σ{sub MUF}. The MBA design is challenging for nuclear facilities managing large amount of bulk (instead of items) SNM because of potentially large MUF and σ{sub MUF}. A concept and guidance of MBA design for uranium fueled HTGRs has been discussed. In this study, a MBA design is presented for the transuranic fuel cycle employing HTGRs. The MUF and the σ{sub MUF} is calculated and evaluated using international target values (ITVs) provided by the IAEA. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Neutronics and computational fluid dynamics analyses for a natural circulation loop of molten fuel salt

A conceptual reactor system of molten fuel salt loop designed for Mo-99 production was analyzed by coupling TRITON neutronics and Fluent computational fluid dynamics (CFD) simulations. The loop system operated in 1 MW thermal power features natural circulation and potential extraction of non-soluble fission products (FP) in fuel salt. The behavior of natural circulation was studied and the power and temperature distributions in equilibrium were obtained. A long-term fuel depletion analysis of the system considering fuel circulation and reprocessing was then performed using an approximate stepwise sequence of simulations. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

CANDLE Burning Reactor with Plutonium Fuel Start-up Core

The CANDLE burning fast reactor concept makes it possible to utilize natural uranium or depleted uranium with high burnup of up to 40% without repossessing of spent fuel from the reactor. Once the reactor reaches equilibrium burning wave condition, it needs only natural or depleted uranium, although it needs some amount of fissile material for the startup. Studies have been performed for the startup core using enriched uranium and absorbers. It is expected that plutonium obtained by reprocessing of spent fuel from light water reactors (LWR) will be an excellent fuel for the startup core. Using this fuel for the startup core will eliminate the need to produce enriched uranium, as well as make it possible to utilize the plutonium from spent fuel effectively as a useful resource. For effective utilization of plutonium as a resource, a startup core that requires only a small amount of plutonium fuel is preferable. The purpose of this study was to clarify the CANDLE burning fast reactor concept using a minimum amount of plutonium fuel from LWR spent fuel for the startup core by analysis with a simplified reactor model. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Aging Processes of Silver Mordenite and Silver Functionalized Aerogel in Dry Air, Humid Air and NO/N{sub 2}

One major problem in the treatment of nuclear-fuel reprocessing off-gas streams is the loss of capacity of the adsorbents exposed to off-gas streams containing air, I{sub 2} vapor, H{sub 2}O vapor, NO and NO{sub 2} over extended periods of time. These aging processes decrease the adsorption capacity of the reduced silver exchanged mordenite (Ag{sup 0}Z) and silver functionalized silica aerogel (Ag{sup 0}-aerogel) columns during the long processing time of iodine adsorption. Significant work has been performed over the past few years at ORNL on Ag{sup 0}Z and Ag{sup 0}-aerogel aging. The Ag{sup 0}Z and Ag{sup 0}-aerogel were aged in 4 different gas streams at 150 deg. C for periods of time up to 6 months. The aging samples were then loaded with I{sub 2} to determine the capacity loss after the aging. The results are summaries in Tab. 1. It was found that both Ag{sup 0}Z and Ag{sup 0}-aerogel had different amounts of capacity loss in different gases. For example, 40% and 22% capacity loss were observed for Ag{sup 0}Z and Ag{sup 0}-aerogel when aged in dry air for 6 months, while 78% and 43% capacity loss were observed when aged in 1% NO{sub 1} for 2 months. In addition, it was found that Ag{sup 0}Z and Ag{sup 0}-aerogel had a different capacity loss when aged at the same conditions. For example, after aged in 2% NO{sub 2} for 2 months, Ag{sup 0}Z lost 30% iodine capacity, but there was no capacity loss for Ag{sup 0}- aerogel. The differences in capacity loss should be due to different aging processes occurred when Ag{sup 0}Z and Ag{sup 0}- aerogel were exposed to different gases. The mechanisms of the aging processes have not been determined and need to be studied with proper chemical and physical characterization methods. Towards this objective, experimental work is proposed and on-going to obtain data to determine the mechanisms of the aging processes of Ag{sup 0}Z and Ag{sup 0}-aerogel in different aging gas streams. In addition, models will be developed to describe the deactivation process of the adsorbents when exposed to off gases. The deactivation model will be included in the transport models for predicting the adsorption processes over these long-time periods.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Nuclear Fuel Cycle Simulator as a Means to Model a Nuclear Hybrid Energy System

With a growing concern for how greenhouse gas (GHG) emissions impact climate change, there is a clear value in minimizing reliance on fossil fuels. Fluctuating renewable energy sources continue to penetrate the grid at higher levels. One means of ensuring a reliable electricity supply, while maintaining low overall GHG emissions, is to couple a nuclear power plant to the fluctuating renewable electricity source. Nuclear hybrid energy systems (NHES) focus on combining renewable and nuclear generation with industrial processes to ensure reliable energy generation as well as profitability of the system as a whole. This paper analyzes the current state of NHES models, and suggests how a nuclear fuel cycle simulator (NFCS) could be used to quickly model a nuclear hybrid energy system to address long term policy concerns. NFCSs are used to model the nuclear fuel cycle, which consists of uranium mining, enrichment, the reactor, reprocessing, and long term storage of waste. The simulations can span over hundreds of years to demonstrate different approaches to managing costs, waste disposal, and proliferation concerns. Current NHES models, on the other hand, focus on shorter time periods, primarily on a daily to weekly scale, and answer questions surrounding feasibility and costs for both normal and extreme scenarios. First, this paper will discuss the different criterion that are important to modeling a NHES. Then, we will break down the characteristics of a NFCS, analyzing how the characteristics of NHES and the fuel cycle overlap. After discussing the current modeling approaches for NHES and NFCS, this paper will analyze the benefits of using a NFCS to model a NHES. This paper will conclude that a NFCS, if appropriately augmented to account for the dynamic aspects of a NHES, can produce a flexible, high level NHES model quickly.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Electrochemical Behavior of Lanthanum in Molten LiF-NaF-KF 'Flinak' Eutectic Salt

In order to develop the Fluoride High- Temperature Reactor, an emerging Gen-IV reactor concept, a primary coolant cleanup system must be developed to allow online reprocessing of the primary molten fluoride salt. This research is focused on the determination of electrochemical properties of rare earths (REs) in molten fluoride salt related to the FHR. The FHR utilizes two fluoride salts. These fluoride salts are LiF-NaF-KF (46.5-11.5-42 mol%), melting point 454 deg. C, also known as 'Flinak', and LiFBeF{sub 2} (33-67 mol%), melting point 459 deg. C, also known as 'Flibe'. Flibe is selected as the primary coolant salt due to its low neutron cross-section (Li{sup 7}-enriched), superior heat transfer properties, and low melting point. However, due to the beryllium content in Flibe, it is highly toxic to humans and Flinak is preferred for the secondary and tertiary coolant salts. The work presented here is specifically focused on the behavior of lanthanum trifluoride (LaF{sub 3}) in molten Flinak. Furthermore, updated data and analysis regarding the electrochemical behavior of dysprosium trifluoride (DyF{sub 3}) is also presented. Previous work related to this topic was presented at the 2016 ANS winter conference. The goal of this research is to develop a contaminant cleanup system for the primary loop of the FHR.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

A study on the criticality safety for the direct disposal of used nuclear fuel in Japan: Application of burnup credit to the criticality safety evaluation for the disposal canister - 14049

Japan has recently started the technical development of the direct disposal of used nuclear fuel (UNF) to prepare various disposal options. Unlike the disposal of radioactive waste from UNF reprocessing, criticality safety is important here because of the presence of a certain amount of fissile nuclides in UNF. This paper provides an outline of the research to be addressed in this field and the relevant studies in Japan. It especially includes the first result of a criticality safety evaluation of a disposal canister model adopting burnup credit. In this evaluation, it was assumed that a steel disposal canister loaded with fuel assemblies was surrounded by a clay buffer material, and the fuel assemblies and the disposal canister were intact. Conditions representative of discharged pressurized-water reactor (PWR) fuel were applied in this study - initial U-235 enrichment of 4.5 wt% and burnup of 45 GWd/t. The isotopic composition of the burnt fuel was evaluated using SWAT3.1 code. In the depletion calculation, conservative parameters were applied within the typical PWR operation range. The effective neutron multiplication factor (k{sub eff}) was then calculated using a continuous-energy Monte Carlo code MVP-2.0, adopting the JENDL-4.0 nuclear data library. Uncertainty in k{sub eff} due to depletion calculation errors was evaluated using the correction factors of isotopic composition, which were obtained via multiple comparisons between the calculated and measured isotopic concentrations of irradiated fuels. In addition, the effects of axial burnup profile and horizontal burnup gradient on k{sub eff} were evaluated using the open burnup distribution database. The k{sub eff} including the uncertainties and conservatism was found to be below 0.95 for both actinide-only and actinide and fission product considerations when the fuel assemblies and the disposal canister were assumed to be intact. (authors)

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