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At least 343 records · Page 19

A simple structure of all circular-air-holes photonic crystal fiber for achieving high birefringence and low confinement loss

We propose a simple structure of photonic crystal fibers (PCFs) with high birefringence and low confinement loss based on one rectangular centric ring of smaller circular air holes (CAHs) in the fiber core, and three rings of larger CAHs in the fiber cladding. This simple geometry (using all CAHs with two different air hole sizes) is capable of achieving a flexible control of the birefringence, B = 5.501 × 10{sup −3}, and ultra-low confinement loss, 7.30 × 10{sup −5 }dB/km, at an excitation wavelength of λ = 1550 nm. The birefringence value is ∼5.0 times greater than that obtained for conventional CAH PCF. This simple structure has the added advantage from the view point of easy fabrication, robustness, and cost. A full-vector finite element method combined with anisotropic perfectly matched layers was used to analyze the various fiber structures. We have analyzed four cases of CAH PCFs, focusing on the core asymmetry design as opposed to the conventional approach of CAHs or elliptical air holes on the cladding and core. The robustness against manufacturing inaccuracies of the proposed structure has also been further investigated in this work.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗

AREVA NP MR Cladding benefits for proposed US NRC RIA and LOCA requirements

M5 is the reference alloy of AREVA NP for PWR cladding applications. It is a zirconium-based alloy containing 1% niobium, no tin, and well-controlled oxygen, iron, and sulfur contents. It has a fully recrystallized microstructure. The absence of tin in M5 results in a very low corrosion rate and hydrogen pick-up during normal operation, even at high burn-up and in high duty irradiation conditions. Due to this low hydrogen pick-up under irradiation (typically less than 100 ppm, End of Life), M5 retains a very good strength and an exceptional residual ductility during and after the accident. This allows utilities implementing M5 to simplify the safety analysis and to demonstrate significant margins regarding fuel related criteria. Simplified methods are easier for plants to implement and maintain for future plant operations and core design management. This paper specifically presents 3 examples on the safety improvement brought by M5 cladding. First, during a reactivity-induced accident (RIA), the properties of M5 are such that bounding analyses for future cycles are significantly simplified, in particular by demonstrating that fuel dispersal will not occur. The major aspects of the AREVA NP RIA methodology are presented. A second focus is made on the low hydrogen embrittlement of M5 and on its high ductility after quenching during a loss-of-coolant accident (LOCA). In a third part, the breakaway oxidation testing program currently under development in Paimboeuf (France) is described, including details about the testing devices and qualification. The tests have demonstrated that M5 cladding does not have breakaway at least up to 5000 s oxidation duration at 1000 C. degrees, which is much higher than the LOCA duration.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Thermo-Mechanical Analysis of SBLOCA with Loss of One Safety Injection System Using SCDAP/RELAP5

Design Extension Condition (DEC) for nuclear power plants has been being deliberated to bring it within a safety regulation boundary, following the 2011 Fukushima Daiichi accident. Moreover, a new regulation on the LOCA accident analysis designated as 10CFR50.45(c) has been proposed based on a concern of the non-conservatism of current criterion neglecting fuel burn-up effect on the cladding embrittlement. Therefore, a more comprehensive and realistic simulation of fuel rod behavior is required to cope with the stringent environment under LOCA accidents. A SCDAP/RELAP5 system analysis code has been used mainly on the prediction of severe accident phenomena. It has a set of mechanistic models for the fuel deformation, which plays an important role particularly in the early phase progression of the fuel melting. However, limited work has been done focusing on the models only to the fuel failure estimation through the DBA or candidates of DEC. Therefore, some analyses were performed here with a burnup dependent core model to understand the capability of SCDAP/RELAP5 to predict the thermo-mechanical behavior of the fuel cladding such as ballooning, burst and external/internal oxidation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Evolution of pores in the fuel cladding of the electrogenerating channel

The results of reactor tests of carbonitride fuel in a monocrystalline cladding from a molybdenum-based alloy that were carried out earlier on an experimental setup Ya-82 for 8 to 300 h at a temperature of ∼1500°C can be used in order to prove the operational reliability of fuel elements in the design of a megawatt nuclear power plant for a spacecraft. A raster image of the surface of a sample of cladding shows that the interfaces between the layers are decorated by pores. This result is explained in this work by the theory of coalescence. The mechanisms responsible for the evolution of pores taking place at the parameters of a Ya-82 setup are considered. The effect of decoration by pores of the interfaces between the layers of a sample of cladding of the electrogenerating channel by carrying out reactor tests is explained. The dependence of the average radius of pores on the duration of the experiment is obtained. An evaluation of the average sizes of pores arising under the conditions of the experiment gives a value of ∼2 μm, which is in agreement with the experimental data. A computational study of swelling of the cladding material in the process of irradiation is performed. Predictive estimates of the behavior of the porous system and swelling of the cladding material for a megawatt class nuclear power plant are made.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

PWR Core Analysis for Cycle Extension and Uprates with LEU+ Accident Tolerant Fuel and 80 GWd/Tonne Burnup Limit

The U.S. Nuclear Regulatory Commission has recently drafted a rule enabling fuel burnup increase in light water reactors up to 80 GWd/t. In conjunction with use of fuel enrichment up to 10%, and accident tolerant fuel (ATF), this is anticipated to facilitate 24-month cycles in PWRs, along with further power uprates. In this paper, PWR core analysis is performed for 20% increased PWR power output along with cycle extension up to 24 months, in combination with use of chromia-doped fuel and chromium-coated clad, considered to be the most near-term ATF concepts. In combination, these lead to challenging conditions with a core average discharge burnup of up to ~74 GWd/t, challenging even the 80 GWd/t burnup limit. Analysis is performed using the 2-step method with POLARIS (within SCALE) used for the lattice calculations and PARCS for the core calculations. Core designs are first baselined for current operating conditions (LEU, 62 GWd/t discharge burnup limit) and then derived that meet cycle constraints on power distribution and the updated lead pin discharge burnup limit while maintaining at least two batches of fuel in the core. Gadolina loadings in fuel pins of up to 8% are used, with enrichment zoning both within the core and, to a limited extent, within assemblies. Here, doped fuel with coated cladding can utilize the same core designs as the reference UOX cores, exhibiting slightly lower burnup due to higher fuel density, which also offsets the slight reactivity penalty from the doping and coating. For the analysis performed here, doped fuel enabled a core with 24-month cycle and 20% uprate to stay within the 80 GWd/t lead pin discharge burnup limit.

LEU+↗

The influence of cladding on fission gas release from irradiated U-Mo monolithic fuel

The monolithic uranium-molybdenum (U-Mo) alloy has been proposed as a fuel design capable of converting the world’s highest power research reactors from use of high enriched uranium to low enriched uranium. However, a zirconium (Zr) diffusion barrier must be used to eliminate interactions that form during fabrication and are enhanced during irradiation between the U-Mo monolith and aluminum alloy 6061 (AA6061) cladding. One aspect of fuel development and qualification is to demonstrate appropriate understanding of the extent of fission product release from the fuel under anticipated service environments. An exothermic reaction has previously been observed between the AA6061 cladding and Zr diffusion layer. In this paper, two fuel segments with different irradiation history were subjected to specified thermal profiles under a controlled atmosphere using a thermogravimetric/differential thermal analyzer coupled with a mass spectrometer inside a hot cell. Samples from each segment were tested with cladding and without cladding to investigate the effect, if any, that the exothermic reaction has on fission gas release mechanisms. Measurements revealed there is an instantaneous effect of the cladding/Zr exothermic reaction, but not necessarily a cumulative effect above approximately 973 K (700 oC). The mechanisms responsible for fission gas release events are discussed.

cl↗

Assessment of SiC-CMC compatibility with components of actinide fuel systems

Silicon carbide (SiC) fiber reinforced SiC ceramic matrix composites (CMC) represent an alternative cladding material for accident tolerant fuel light water reactor (LWR) applications, however, data and technology gaps exist that need to be understood before SiC-CMCs can be considered. One technology gap is the understanding of fuel-clad chemical interactions (FCCI) in SiC-CMC clad LWR systems. A review of SiC interactions with relevant elements to the LWR fuel system is discussed. Insights are drawn from research focusing on the use of SiC for high temperature electronics applications and high temperature structural materials applications as well as insights from tristructural-isotropic (TRISO) coated particle fuel development, where SiC is a critical component of the TRISO design. The interactions discussed in this review are framed relative to the expected conditions for SiC-CMC cladding application. Furthermore, the general observation is that limited risk of FCCI exists under normal operating conditions for SiC-CMC cladding systems in LWR applications.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

A model to assess Zircaloy’s mechanical property changes following a transient beyond critical heat flux

Maintaining the integrity of nuclear fuel rods is essential for ensuring public health and safety in nuclear power generation. During reactor operation, this integrity is confirmed by demonstrating compliance with established regulatory acceptance criteria. For moderate-frequency events, such as limiting transients and anticipated operational occurrences (AOOs), the current fuel integrity criterion is based on preventing boiling transition. This criterion assumes that prevention of boiling transition will prevent excessive cladding heating and, thus, fuel failure during normal operations. While conservative, this approach places significant constraints on core design, fuel cycle economics, and a plant’s ability to perform major power uprates, leading to suboptimal fuel utilization and inefficient carbon-free energy production. A more efficient approach could be achieved by revising the failure criterion to a material-specific limit rather than strictly preventing the boiling transition, since boiling transition per se is not a cause of fuel cladding failure. Here, as a result, a new licensing framework based on material properties, termed time-at-temperature (t@T), is needed. This approach would allow for brief periods of post–critical heat flux operation during an AOO without compromising safety. Implementing the t@T licensing strategy requires a robust technical foundation in material properties, which must be established through comprehensive data collection on both unirradiated and irradiated fuel and cladding materials. This foundation would enable the development of a safety basis that ensures safe operation while providing greater flexibility and efficiency for reactor operation. This paper documents a thorough review of the available data to establish a baseline knowledge that can inform the development of cladding mechanical models, as well as identify experimental data gaps that need to be addressed in future research. Machine learning and data informatics were utilized to extract the importance of parameters on the t@T parameter. Industry tools were used to perform baseline analyses to define the relevant transient conditions for data analysis. The subsequent review successfully identified applicable experimental data, as well as sufficient data to evaluate changes in cladding mechanical properties following an AOO transient. Rather than developing new models, this work coupled existing irradiation annealing and recrystallization models to calculate changes in hardness, yield stress, and ultimate tensile stress following an AOO event. The findings from this review were summarized to highlight the experimental data needs required to fill remaining gaps and support the development of future t@T licensing methodologies.

Cladding performance↗

Evolution of Microstructure and Surface Characteristics of FeCrAl alloys when Subjected to Flow Boiling Testing

FeCrAl alloys are candidate materials for manufacturing accident-tolerant fuel (ATF) cladding intended for light water reactors to increase fuel reliability and safety during design-basis and beyond-design-basis accident scenarios. The evolution of the materials' surface characteristics, microstructure, and mechanical properties when exposed to the Critical Heat Flux (CHF) in flow boiling testing is crucial for safety analysis while providing insights into their thermal-hydraulic performance in nuclear reactors. After CHF, the surface chemistry of two FeCrAl alloys, APMT and C26M, was studied to understand their evolution at the early stage of high-temperature excursions in short time periods. The results indicated a thin layer composed of oxides and hydroxides of Al, Cr, and Fe with varying proportions at different depths in the layer, as indicated by X-ray photoelectron spectroscopy (XPS) and depth profiling. The cross-sections prepared by focused ion beam (FIB) revealed the growth of an oxide layer, in the range of 90-180 nm thick, on the alloys' surfaces. The evolution of the materials' surface chemistry also led to a noticeable post CHF excursion increase in their wettability, with a slight increase in roughness. Further, the investigation of the materials' mechanical properties indicated a modest increase in hardness by 10-15% as well as an increase in their yield strength, as evidenced by the microindentation and ring compression tests conducted before and after CHF testing. Scanning electron Microscopy (SEM) and X-ray diffraction (XRD) were used to investigate microstructural features of the materials and their changes after CHF treatment.

36 MATERIALS SCIENCE↗

Highly non-linear solid core photonic crystal fiber with one nano hole

The numerical study of newly designed solid core photonic crystal fiber (SCPCF) having three hexagonal air hole rings in cladding region and one small nano hole at the center are presented. By using full vectorial finite element method (FV-FEM), we analyses the optical properties like effective area, nonlinearity and confinement loss of the proposed PCF. Results show that the change in core diameter controls the effective area, nonlinearity and confinement loss. A low effective area (3.34 µm{sup 2}), high nonlinearity (36.34 W{sup −1}km{sup −1}) and low confinement loss (0.00106 dB/km) are achieved at the communication wavelength 1.55 µm for the SCPCF having core air hole diameter 0.10 µm, cladding air holes diameter 1.00 µm and pitch 2.50 µm. This type of PCF is very useful in non-linear applications such as supercontinuum generation, four wave mixing, second harmonic generation etc.

36 MATERIALS SCIENCE↗

The Impact of the Halden Reactor Project on the Development of the Bison Fuel Performance Code

The Bison fuel performance code has been under development at Idaho National Laboratory by various programs within the U.S. Department of Energy since 2009. As part of any fuel performance code development, validation is necessary. To this end, several Halden experiments have been used. Early on, the focus was on validating the thermal mechanical behaviour, including pellet cladding mechanical interaction and fission gas release models implemented in Bison. Support for design-basis accident capabilities began in approximately 2014. Also in 2014, the development of models for accident tolerant fuel (ATF) concepts also began. This includes dopedUO2 fuel for which Halden has some experiments with temperature and fission gas release measurements. The Bison team has been fortunate to have two senior developers be secondees to Halden for a year: the late Dr. Giovanni Pastore from May 2015 to April 2016 and the now retired Dr. Richard Williamson in 2018. The paper will provide examples of Bison validation using Halden data for standard UO2 fuel under normal operation and loss-of-coolant accident condition, and ATF concepts. A tribute to the legacy that both the late Giovanni and retired Richard left on the Bison code is provided.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Extension and Demonstration of NEAMS Multiphysics Tools to Lead-Cooled, Sodium-Cooled, and Molten Salt Fast Reactor Applications

The SHARP toolkit is a high-fidelity reactor simulation tool developed under the U.S. Department of Energy, Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. SHARP toolkit is comprised of the neutronics module PROTEUS thermal hydraulics module Nek5000, and structural mechanics module Diablo. During FY17 and FY18, the PROTEUS and Nek5000 components of SHARP were applied to solve challenging sodium-cooled fast reactor (SFR) problems. In particular, selected hot channel factors (HCF) for a prototype metal-fueled SFR design (the AFR-100) were analyzed in high fidelity, and the “SHARP zooming capability” for SFRs was developed and demonstrated to reduce computational expense for full core problems in cases where detailed data is needed in selected fuel assemblies. After the previous success applying SHARP to challenging SFR problems, the focus in FY19 expanded to additional fast reactor applications: lead cooled fast reactors (LFR), sodium cooled fast reactors (SFR), and fast molten salt reactors (MSR). The specific technical tasks were (1) assessment of hot channel factors for LFR, for which no data currently exists, (2) implementation of a gamma transport capability in the PROTEUS solvers to extend the zooming capability to non-fueled assemblies, (3) demonstration of zooming capability in assemblies of the Versatile Test Reactor (VTR), and (4) development of a new coupled capability for simulating fast MSRs, specifically modeling precursor transport inside and outside the core. The progress on tasks (1) and (3) are discussed primarily in this document. Tasks (2) and (4) are briefly mentioned in this document, and the interested reader is referred to separate reports for detailed information. First-of-a-kind hot channel factor (HCF) estimation for LFR with high fidelity codes (PROTEUS/Nek5000) was successfully demonstrated in this study. Selected HCF were computed and compared with SFR data (AFR-100, EBR-II). The findings confirm that different reactor types, design parameters and uncertainties lead to different HCFs. Careful estimation of HCF for a specific design is necessary to obtain appropriate HCFs. In addition to improvement in HCF accuracy, high fidelity tools generate data to help the designer better understand the mechanism of the impact from these uncertainties. For example, the impact of cladding thickness manufacturing tolerance resulted in non-intuitive effects in the corner pins of the LFR assembly. This procedure of computing HCF using high fidelity models shows promise and flexibility for being repeated for any arbitrary reactor of choice, since these tools are targeted to solve many reactor types and geometries. Progress was made towards extending the previously demonstrated SHARP zooming capability to non-fueled SFR assemblies. In particular, a gamma transport capability was implemented in both high fidelity PROTEUS solvers in order to accurately account for heat deposition caused by gamma particles, which accounts for ~10% of total core power. Neutronics verification cases were carried out for a candidate Versatile Test Reactor (VTR) design using the new gamma transport capability in PROTEUS. Comparisons were made with continuous energy MCNP calculations and shown to agree well. The models for the full core design with heterogeneous control and fuel assemblies is in progress and will be completed in FY20. In the fast molten salt reactor (MSR) domain, a new capability was developed to model the flow of neutron precursors both inside and outside of the core, which impacts the effective delayed neutron fraction of the system. The PROTEUS-SN and PROTEUS-NODAL solvers were coupled to Nek5000 in order to simulate precursor flow. The workflows were demonstrated on test problems. The fast MSR modeling capability was supported under two other work packages, and comprehensive details are included in a separate report.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Spent Nuclear Fuel Transportation: Rail Conveyance Transmissibility

Equipos Nucleares, S.A. (ENSA) and the US Department of Energy (DOE) are preparing a full scale Spent Nuclear Fuel (SNF) transportation test. This transportation testing will include road, rail, coastal, and trans-Atlantic shipments. The test campaign will use a full scale commercial transportation package and cradle. The package will be loaded with two instrumented fuel assemblies to measure strains at cladding locations and accelerations on the fuel assemblies, and thirty “dummy” assemblies. This testing is designed to closely match an actual SNF shipment. Accelerometers will also be used at various locations throughout the full conveyance system (i.e. rail car/truck, cask, transport cradle, and basket) to study the transmission of loads through the system and to provide validation for numerical models. Previous testing and modeling work has shown how the structural transmissibility of the transport system can affect the magnitude of these loads, and the importance of modeling all aspects of the transport system (i.e. rail car/truck, transport cradle, cask, basket, and fuel) (Ref 1, 2). This paper describes preliminary models that were constructed to estimate load transmission during rail transport, from the bottom of the cradle to an individual fuel rod within the package. The modeling studies in this paper evaluate the system response to postulated shock pulses and random vibration loads. These models describe the transmissibility of the conveyance, and demonstrate how loads can be amplified or attenuated as they are transmitted through the structure. This is done by coupling the rail vehicle dynamics code NUCARS, to the general finite element modeling code ANSYS, and the explicit dynamics code LS-DYNA. Models such as the ones presented herein will be used during the test campaign to help analyze and evaluate the test data as it is collected.

Cladding, Spent Fuel and Radiactive Waste, Used Fu↗

Correlation between Grain Size Distribution and Grain Boundary Character in Polycrystalline Uranium Dioxide

Thermo-mechanical behavior of oxide nuclear fuels is extremely important to overall fuel element performance. Critical to the understanding of Pellet-Cladding Mechanical Interactions (PCMI) are elasticity, plasticity and creep, as this interaction with the cladding can lead to fuel rod fracture and in turn affect its thermal performance significantly. In addition, sub-grain scale mechanical behavior, e.g. elastic anisotropy, as well as dislocation-driven plasticity and creep, can also play a significant role on microstructure evolution of the fuel elements and their cladding, failure of which would severely reduce the useful service life of the elements and allow the release of fission products out into the working fluid of the reactor. Therefore, careful measurements of mechanical properties are key to validate robust fuel performance codes able to predict this PCMI behavior from inputs at the micro-scale, e.g., MARMOT, and to quantify its effects on other aspects of fuel behavior. In uranium dioxide (UO{sub 2}), the thermomechanical response at the meso-scale depends strongly on crystallography of individual grains, and available datasets are far from complete in terms of temperature, stress, and stoichiometry in parameter space. Therefore, details of the fuel microstructure, such as porosity distribution and location, as well as grain boundary (GB) character, distribution, and location are needed to create accurate multiscale models to predict the fuel's evolution in-pile, and in turn aid in the design and development of more efficient fuels that can last longer cycles.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Monitoring the Possible Effects of Corroding Fuel Cladding in Nuclear Ponds: a Modified ROV for Storage Ponds - 20144

Nuclear ponds are used worldwide to store radioactive nuclear material, often for extended periods of time. The pond water conditions need to be closely monitored so that mitigation strategies can be put in place should any leakage of radioactive material occur. This paper describes the selection, development and on-site deployment of a bespoke system whose design was based on of-the-shelf components. The system is able to identify, locate and differentiate the leakage of radioactive material stored in nuclear ponds from the background activity of the water in the pond. The selection of an appropriate detector and monitoring geometry was supported by MCNP modelling which allowed a high degree of confidence in the proposed design. The development of the system allowed a real-time indication of the canister integrity, without incurring the increased operator doses from manual sampling operations that were previously used on site. The benefits of the developed system include the fact that the system is lightweight, easily deployable, retrievable, decontaminable and results in negligible waste products. Moreover, it is able to analyse the samples in-situ, in this way reducing the time needed to manually collect and send the samples to a laboratory for lengthy analysis. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

In-Pile Loss of Coolant Accident Testing at TREAT

Several hot-cell based LOCA test apparatuses still exist in various institutions across the world. These facilities typically surround rods with infrared electrical heaters to simulate LOCA conditions on the cladding. Despite its utility for cladding-focused testing, the furnace-based external heating method creates an inverted temperature gradient in fuel pellets compared to a true LOCA where radioisotope decay heat generates internal heat. Capturing the correct temperature gradient is essential in simulating the thermomechanical stress in fuel pellets, and hence a crucial design parameter investigating FFRD during a postulated LOCA. This thermomechanical stress state, for which the evolution of fuel pellet centerline and cladding temperatures is a good figure of merit, was the primary parameter investigated during conception of the LOCA-SERTTA device.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Progress in Fast Modular Reactor Conceptual Design

The Fast Modular Reactor (FMR) is a 100-MW(thermal) gas-cooled fast reactor being developed by General Atomics Electromagnetic System with the goal of developing a FMR for flexible and dispatchable power to the U.S. electricity market in the mid-2030s. The conceptual design aims to develop and verify simplified design features. These include an inert helium gas coolant, pellet-loaded fuel rods, installations with air cooling as ultimate heat sink, and small and passive heat removal systems. The goal is to ensure the development of a safe, maintainable, cost-effective, and distributed nuclear energy-generating station. The baseline technologies selected to achieve this goal are a helium coolant that is an inert gas with no chemical reaction with structural components, not activated, single phase, enabling high-temperature operation and a high thermal efficiency Brayton cycle; conventional uranium dioxide (UO 2 ) fuel, which is the most widely used and well-known fuel material, capable of high burnup (100 MWd/kg) and a long fuel life; and silicon carbide composite (SiGA®) cladding and internal structures that are chemically inert in the helium environment, exceptionally radiation tolerant, and being derisked by accident tolerant fuel technology development. Further, the reactor was specifically designed with passive safety features, including high-temperature in-core materials and a reactor vessel cooling system consisting of cooling panels of naturally circulating water. The passive safety of the core was confirmed for the depressurized loss-of–forced cooling accident, which showed the peak cladding temperature at ~1600°C during the transient, which is below the current design limit of 1800°C. The conceptual design of the FMR has been conducted for the reactor system, vessel system, generator and turbomachine, instrumentation and control, residual heat removal system, plant service system, and containment, as well as pre-application licensing documents.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Technology Development for Dry Storage of Aluminum-Clad Spent Nuclear Fuel - 20490

A candidate disposition pathway for the > 13 MTHM of aluminum-clad spent nuclear fuel (ASNF), owned and managed by the U.S. Department of Energy, is the drying and placement of the SNF into sealed-canister dry storage, with the ASNF-in- canisters 'road-ready' for transportation to and final direct disposal in a repository waste package. Technical information gaps in fuel drying, and fuel dry storage behavior, have previously challenged the declaration of technology readiness for drying and placement of this fuel into the DOE Standard Canister design for > 50 years of safe dry storage. The principal technical information gaps included: i) characterization and thermal dehydration behavior of aluminum (oxy)hydroxide films attendant on the cladding due to film formation during reactor operation and during post-discharge up to long-term wet storage histories; and ii) G-value data to enable estimation of the radiolytic gas generation from the cladding with its (oxy)hydroxide films. Thus, the oxide films on the ASNF challenged the safety of a sealed storage canister with thermal and radiolytic decomposition of the waters on the films that can lead to corrosion, pressurization, and flammability issues. These gaps in the technical information base have largely been closed. This paper discusses the investigations at the Idaho National Laboratory (INL) and the Savannah River National laboratory (SRNL), and outlines the pending technology development work for input to an engineering design to enable a road-ready dry storage system for ASNF. The ASNF inventory considered for road-ready dry storage is stored at the Savannah River Site (SRS) and at the Idaho Nuclear Technology and Engineering Center (INTEC) at the INL. The ASNF inventory in the SRS L Basin is from foreign and domestic research reactors (FRR and DRR), and is diverse in terms of design, irradiation, and post-reactor-discharge storage conditions; these factors yield a range of characteristics of cladding oxide films on ASNF. Mixed aluminum (oxy)hydroxide (boehmite and bayerite/gibbsite) films, non-uniform in thicknesses up to a maximum local thickness bounded by 25 μm, were observed on ASNF materials removed from wet storage in the L Basin and in non-sealed dry storage at the SRS following reactor service and a long-term (up to 40+ years) interim storage history. The ASNF inventory at INTEC, in both wet and vented dry storage, is predominantly from the Advanced Test Reactor (ATR), but it also includes DRR and FRR fuel. To address a profound behavior of these films, radiolysis testing of aluminum specimens with mixed type boehmite/bayerite oxide films was performed to develop basic data on G-values for production of radiolytic hydrogen under dry storage conditions with nominal relative humidity, temperature, and cover gas. Modeling and simulation of canister internal environments with postulated inventories of oxide films provides estimation of the evolution of the conditions of the canister loaded with ASNF. Simultaneous Thermal Analysis (TGA/DSC) of hydrated oxide powders, and laboratory-scale aluminum specimens with a bayerite film (∼10 μm) using TGA methods, inform drying time/temperature conditions to be used for the ASNF. A demonstration project is recommended for Verification and Validation of the drying and storage of the ASNF. Remaining major tasks leading up to the hot demonstration include scale-up radiolysis testing and scale-up drying testing. Engineering design with the information from the technology program will establish the safety basis and enable long-term (> 50 years) dry storage compatibility with ASNF in the DOE Standard Canister pending its transportation to and disposal in a repository. This full capability would show the ASNF-in-canister storage system to be road-ready. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗