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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 1,711 records · Page 95
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Measure Guideline: Deep Energy Enclosure Retrofit for Double-Stud Walls

This Measure Guideline describes a deep energy enclosure retrofit (DEER) solution that provides insulation to the interior of the wall assembly with the use of a double stud wall. The guide describes two approaches to retrofitting the existing the walls: one involving replacement of the existing cladding, and the other that leaves the existing cladding in place. It discusses the design principles related to the use of various insulation types, and provides strategies and procedures for implementing the double stud wall retrofit. It also evaluates important moisture-related and indoor air quality measures that need to be implemented to achieve a durable, high performance wall.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2014

Thermal Analysis of a TREAT Fuel Assembly

The objective of this study was to explore options as to reduce peak cladding temperatures despite an increase in peak fuel temperatures. A 3D thermal-hydraulic model for a single TREAT fuel assembly was benchmarked to reproduce results obtained with previous thermal models developed for a TREAT HEU fuel assembly. In exercising this model, and variants thereof depending on the scope of analysis, various options were explored to reduce the peak cladding temperatures.

Papadias, Dionissios↗
DOE OSTI2015

Test Plan for the Boiling Water Reactor Dry Cask Simulator

The thermal performance of commercial nuclear spent fuel dry storage casks are evaluated through detailed numerical analysis . These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. These same vertical, canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and below-ground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of above-ground and below-ground canistered dry cask systems. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern vertical, canistered dry cask systems. The BWR cask simulator (BCS) has been designed in detail for both the above-ground and below-ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 deg C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the canister. The symmetric single assembly geometry with well-controlled boundary conditions simplifies interpretation of results. Various configurations of outer concentric ducting will be used to mimic conditions for above and below-ground storage configurations of vertical, dry cask systems with canisters. Radial and axial temperature profiles will be measured for a wide range of decay power and helium cask pressures. Of particular interest is the evaluation of the effect of increased helium pressure on allowable heat load and the effect of simulated wind on a simplified below ground vent configuration. While incorporating the best available information, this test plan is subject to changes due to improved understanding from modeling or from as-built deviations to designs. As-built conditions and actual procedures will be documented in the final test report.

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

Role of Defects in Swelling and Creep of Irradiated SiC

Silicon carbide is a promising cladding material because of its high strength and relatively good corrosion resistance. However, SiC is brittle and therefore SiC-based components need to be carefully designed to avoid cracking and failure by fracture. In design of SiC-based composites for nuclear reactor applications it is essential to take into account how mechanical properties are affected by radiation and temperature, or in other words, what strains and stresses develop in this material due to environmental conditions. While thermal strains in SiC can be predicted using classical theories, radiation-induced strains are much less understood. In particular, it is critical to correctly account for radiation swelling and radiation creep, which contribute significantly to dimensional instability of SiC under radiation. Swelling typically increases logarithmically with radiation dose and saturates at relatively low doses (damage levels of a few dpa). Consequently, swelling-induced stresses are likely to develop within a few months of operation of a reactor. Radiation-induced volume swelling in SiC can be as high as 2%, which is significantly higher than the cracking strain of 0.1% in SiC. Swelling-induced strains will lead to enormous stresses and fracture, unless these stresses can be relaxed via some other mechanism. An effective way to achieve stress relaxation is via radiation creep. Although it has been hypothesized that both radiation swelling and radiation creep are driven by formation of defect clusters, existing models for swelling and creep in SiC are limited by the lack of understanding of specific defects that form due to radiation in the range of temperatures relevant to fuel cladding in light water reactors (LWRs) (<1000°C). For example, defects that can be detected with traditional transmission electron microscopy (TEM) techniques account only for 10-45% of the swelling measured in irradiated SiC. Here, we have undertaken an integrated experimental and modeling effort to discover the previously invisible defects in irradiated SiC and to determine the contributions of these defects to radiation swelling. Knowledge of the most stable defect structures and the rate controlling processes during defect evolution is essential for development of predictive models for swelling and creep as a function of temperature and radiation dose. This research has been enabled by state-of-the-art imaging techniques, such as the aberration corrected scanning transmission electron microscopy (STEM) (FEI TITAN) closely coupled with multi-scale models of stable defect clusters and their evolution.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Improving Limit Surface Search Algorithms in RAVEN Using Acceleration Schemes: Level II Milestone

The RAVEN code is becoming a comprehensive tool to perform Probabilistic Risk Assessment (PRA); Uncertainty Quantification (UQ) and Propagation; and Verification and Validation (V&V). The RAVEN code is being developed to support the Risk-Informed Safety Margin Characterization (RISMC) pathway by developing an advanced set of methodologies and algorithms for use in advanced risk analysis. The RISMC approach uses system simulator codes applied to stochastic analysis tools. The fundamental idea behind this coupling approach to perturb (by employing sampling strategies) timing and sequencing of events, internal parameters of the system codes (i.e., uncertain parameters of the physics model) and initial conditions to estimate values ranges and associated probabilities of figures of merit of interest for engineering and safety (e.g. core damage probability, etc.). This approach applied to complex systems such as nuclear power plants requires performing a series of computationally expensive simulation runs. The large computational burden is caused by the large set of (uncertain) parameters characterizing those systems. Consequently, exploring the uncertain/parametric domain, with a good level of confidence, is generally not affordable, considering the limited computational resources that are currently available. In addition, the recent tendency to develop newer tools, characterized by higher accuracy and larger computational resources (if compared with the presently used legacy codes, that have been developed decades ago), has made this issue even more compelling. In order to overcome to these limitations, the strategy for the exploration of the uncertain/parametric space needs to use at best the computational resources focusing the computational effort in those regions of the uncertain/parametric space that are “interesting” (e.g., risk-significant regions of the input space) with respect the targeted Figures Of Merit (FOM): for example, the failure of the system, subject of the analysis. These methodologies are named, in the RAVEN environment, adaptive sampling strategies. These methodologies infer system responses from surrogate models constructed from already existing samples (produced using high fidelity simulations) and suggest the most relevant location (coordinate in the input space) of the next sampling point to be explored in the uncertain/parametric domain. When using those methodologies, it is possible to understand features of the system response with a small number of carefully selected samples. This report focuses on the development and improvement of the limit surface search. The limit surface is an important concept in system reliability analysis. Without going into the details, which will be covered later in the report, the limit surface could be briefly described as an hyper-surface in the system uncertainty/parametric space separating the regions leading to a prescribed system outcome. For example, if the uncertainty/parametric space is the one generated by the reactor power level and the duration of the batteries, the system is a nuclear power plant and the system outcome discriminating variable is the clad failure in a station blackout scenario, then the limit surface separates the combinations of reactor power level and battery duration that lead to clad failure from the ones that do not.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Severe Accident Analysis Of BWR Core Fueled With UO 2 /FeCrAl with Updated Materials and Melt Properties From Experiments

Iron-chromium-aluminum (FeCrAl) alloys are being considered as advanced fuel cladding concepts with enhanced accident tolerance. FeCrAl alloys have slower oxidation kinetics and higher strength compared to Zr-based alloys at high temperatures. FeCrAl could be used for fuel cladding and spacer or mixing vane grids in light water reactors and/or as channel box material in boiling water reactors (BWRs). There is a need to assess the potential gains afforded by the FeCrAl accident tolerant fuel concept over the existing zirconium based materials employed today.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Report on simulation of fission gas and fission product diffusion in UO 2

In UO 2 nuclear fuel, the retention and release of fission gas atoms such as xenon (Xe) are important for nuclear fuel performance by, for example, reducing the fuel thermal conductivity, causing fuel swelling that leads to mechanical interaction with the clad, increasing the plenum pressure and reducing the fuel–clad gap thermal conductivity. We use multi-­scale simulations to determine fission gas diffusion mechanisms as well as the corresponding rates in UO 2 under both intrinsic and irradiation conditions. In addition to Xe and Kr, the fission products Zr, Ru, Ce, Y, La, Sr and Ba have been investigated. Density functional theory (DFT) calculations are used to study formation, binding and migration energies of small clusters of Xe atoms and vacancies. Empirical potential calculations enable us to determine the corresponding entropies and attempt frequencies for migration as well as investigate the properties of large clusters or small fission gas bubbles. A continuum reaction-­diffusion model is developed for Xe and point defects based on the mechanisms and rates obtained from atomistic simulations. Effective fission gas diffusivities are then obtained by solving this set of equations for different chemical and irradiation conditions using the MARMOT phase field code. The predictions are compared to available experimental data. The importance of the large Xe U3O cluster (a Xe atom in a uranium + oxygen vacancy trap site with two bound uranium vacancies) is emphasized, which is a consequence of its high mobility and high binding energy. We find that the Xe U3O cluster gives Xe diffusion coefficients that are higher for intrinsic conditions than under irradiation over a wide range of temperatures. Under irradiation the fast-­moving Xe U3O cluster recombines quickly with irradiation-induced interstitial U ions, while this mechanism is less important for intrinsic conditions. The net result is higher concentration of the Xe U3O cluster for intrinsic conditions than under irradiation. We speculate that differences in the irradiation conditions and their impact on the Xe U3O cluster can explain the wide range of diffusivities reported in experimental studies. However, all vacancy-­mediated mechanisms underestimate the Xe diffusivity compared to the empirical radiation-­enhanced rate used in most fission gas release models. We investigate the possibility that diffusion of small fission gas bubbles or extended Xe-­vacancy clusters may give rise to the observed radiation-­enhanced diffusion coefficient. These studies highlight the importance of U divacancies and an octahedron coordination of uranium vacancies encompassing a Xe fission gas atom. The latter cluster can migrate via a multistep mechanism with a rather low effective barrier, which together with irradiation-induced clusters of uranium vacancies, gives rise to the irradiation-enhanced diffusion coefficient observed in experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Non-destructive Quantitative Phase Analysis and Microstructural Characterization of Zirconium Coated U-10Mo Fuel Foils via Neutron Diffraction

This report uses neutron diffraction to investigate the crystal phase composition of uranium-molybdenum alloy foils (U-10Mo) for the CONVERT MP-1 Reactor Conversion Project, and determines the effect on alpha-uranium contamination following the deposition of a Zr metal diffusion layer by various methods: plasma spray deposition of Zr powders at LANL and hot co-rolling with Zr foils at BWXT. In summary, there is minimal decomposition of the gamma phase U-10Mo foil to alpha phase contamination following both plasma spraying and hot co-rolling. The average unit cell volume, i.e. lattice spacing, of the Zr layer can be mathematically extracted from the diffraction data; co-rolled Zr matches well with literature values of bulk Zr, while plasma sprayed Zr shows a slight increase in the lattice spacing, indicative of interstitial oxygen in the lattice. Neutron diffraction is a beneficial alternative to conventional methods of phase composition, i.e. x ray diffraction (XRD) and destructive metallography. XRD has minimal penetration depth in high atomic number materials, particularly uranium, and can only probe the first few microns of the fuel plate; neutrons pass completely through the foil, allowing for bulk analysis of the foil composition and no issues with addition of cladding layers, as in the final, aluminum-clad reactor fuel plates. Destructive metallography requires skilled technicians, cutting of the foil into small sections, hazardous etching conditions, long polishing and microscopy times, etc.; the neutron diffraction system has an automated sample loader and can fit larger foils, so there is minimal analysis preparation; the total spectrum acquisition time is ~ 1 hour per sample. The neutron diffraction results are limited by spectra refinement/calculation times and the availability of the neutron beam source. In the case of LANSCE at Los Alamos, the beam operates ~50% of the year. Following the lessons learned from these preliminary results, optimizations to the process and analysis can be made, and neutron diffraction can become a viable and efficient technique for gamma/alpha phase composition determination for nuclear fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Report on fundamental modeling of irradiation-induced swelling and creep in FeCrAl alloys

In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromium-aluminum (FeCrAl) alloys due to much slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high temperature steam, making accident mitigation more likely. As a continuation of the development for these alloys, the material response must be demonstrated to provide suitable radiation stability, in order to ensure that there will not be significant dimensional changes (e.g., swelling), as well as quantifying the radiation hardening and radiation creep behavior. In this report, we describe the use of cluster dynamics modeling to evaluate the defect physics and damage accumulation behavior of FeCrAl alloys subjected to neutron irradiation, with a particular focus on irradiation-induced swelling and defect fluxes to dislocations that are required to model irradiation creep behavior.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Examination of Hybrid Metal Coatings for Mitigation of Fission Product Release and Corrosion Protection of LWR SiC/SiC

There is a need to increase the safety margins of current and future light water reactors (LWRs) due to the unfortunate events at Fukushima Daiichi Nuclear Plant. Safety is crucial to restore public confidence in nuclear energy, acknowledged as an economical, high-­density energy solution to climate change. The development of accident-­tolerant fuel (ATF) concepts is crucial to this endeavor. The objective of ATF is to delay the consequences of accident progression, being inset in high temperature steam and maintaining high thermomechanical strength for radionuclide retention. The use of advanced SiCf-­SiC composite as a substitute for zircaloy-­based cladding is being considered. However, at normal operations, SiC is vulnerable to the reactor coolant and may corrode at an unacceptable rate. As a ceramic-­matrix composite material, it is likely to undergo microcracking operation, which may compromise the ability to contain gaseous fission products. A proposed solution to both issues is the application of mitigation coatings for use in normal operations. At Oak Ridge National Laboratory (ORNL), three coating technologies have been investigated with industry collaborators and vendors. These are electrochemical deposition, cathodic arc physical vapor deposition (PVD hereafter) and vacuum plasma spray (VPS). The objective of this document is to summarize these processing technologies, the resultant as-­processed microstructures and properties of the coatings. In all processes, substrate constraint resulted in substantial tensile stresses within the coating layer. Each technology must mitigate this tensile stress. Electrochemical coatings use chromium as the coolant facing material, and are deposited on a nickel or carbon “bond coat”. This is economical but suffers microcracking in the chromium layer. PVD-­based coatings use chromium and titanium in both metallic form and nitrides, and can be deposited defense-­in-­depth as multilayers. This vapor method eliminated tensile stress during processing and coatings were up to ~30 μm thick without microcracking. VPS produced coatings based on Zircaloy-2, which has a proven reactor-­compatibility. The tensile stresses appearred to be partially mitigated by annealing. Analysis showed that VPS coatings required further optimizations to prevent adverse reactions with the substrate and need a minimum thickness of ~50 μm. In addition, development of coatings are constrained by neutronic depletion analysis, which clearly indicated enrichment as an issue if the coating is too thick. Based on the present work, the cathodic arc PVD technology was considered ready for the extensive testing and evaluation on cladding materials due to their ability to mitigate the excessive tensile stresses and the reasonable coating quality achieved. The VPS Zircaloy-­2 coating technology required additional development toward mitigation of issues related to the substrate reaction and porosity. In the future, PVD and VPS will have be scaled upon successful development and demonstration. Electrochemical coatings, which are proven scalability, currently require development to mitigate issues related to the tensile stress after deposition.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

The SAS4A/SASSYS-1 Safety Analysis Code System, Version 5

The SAS4A/SASSYS-1 computer code is developed by Argonne National Laboratory for thermal, hydraulic, and neutronic analysis of power and flow transients in liquidmetal- cooled nuclear reactors (LMRs). SAS4A was developed to analyze severe core disruption accidents with coolant boiling and fuel melting and relocation, initiated by a very low probability coincidence of an accident precursor and failure of one or more safety systems. SASSYS-1, originally developed to address loss-of-decay-heat-removal accidents, has evolved into a tool for margin assessment in design basis accident (DBA) analysis and for consequence assessment in beyond-design-basis accident (BDBA) analysis. SAS4A contains detailed, mechanistic models of transient thermal, hydraulic, neutronic, and mechanical phenomena to describe the response of the reactor core, its coolant, fuel elements, and structural members to accident conditions. The core channel models in SAS4A provide the capability to analyze the initial phase of core disruptive accidents, through coolant heat-up and boiling, fuel element failure, and fuel melting and relocation. Originally developed to analyze oxide fuel clad with stainless steel, the models in SAS4A have been extended and specialized to metallic fuel with advanced alloy cladding. SASSYS-1 provides the capability to perform a detailed thermal/hydraulic simulation of the primary and secondary sodium coolant circuits and the balance-ofplant steam/water circuit. These sodium and steam circuit models include component models for heat exchangers, pumps, valves, turbines, and condensers, and thermal/hydraulic models of pipes and plena. SASSYS-1 also contains a plant protection and control system modeling capability, which provides digital representations of reactor, pump, and valve controllers and their response to input signal changes.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2017

Evaluation of a Zirconium Recycle Scrubber System

A hot-cell demonstration of the zirconium recycle process is planned as part of the Materials Recovery and Waste Forms Development (MRWFD) campaign. The process treats Zircaloy® cladding recovered from used nuclear fuel with chlorine gas to recover the zirconium as volatile ZrCl4. This releases radioactive tritium trapped in the alloy, converting it to volatile tritium chloride (TCl). To meet regulatory requirements governing radioactive emissions from nuclear fuel treatment operations, the capture and retention of a portion of this TCl may be required prior to discharge of the off-gas stream to the environment. In addition to demonstrating tritium removal from a synthetic zirconium recycle off-gas stream, the recovery and quantification of tritium may refine estimates of the amount of tritium present in the Zircaloy cladding of used nuclear fuel. To support these objectives, a bubbler-type scrubber was fabricated to remove the TCl from the zirconium recycle off-gas stream. The scrubber was fabricated from glass and polymer components that are resistant to chlorine and hydrochloric acid solutions. Because of concerns that the scrubber efficiency is not quantitative, tests were performed using DCl as a stand-in to experimentally measure the scrubbing efficiency of this unit. Scrubbing efficiency was ~108% ± 3% with water as the scrubber solution. Variations were noted when 1 M NaOH scrub solution was used, values ranged from 64% to 130%. The reason for the variations is not known. It is recommended that the equipment be operated with water as the scrubbing solution. Scrubbing efficiency is estimated at 100%.

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

Development of New Heats of Advanced Ferritic/Martensitic Alloys

The Fuel Cycle Research and Development program is investigating methods of transmuting minor actinides in various fuel cycle options. To achieve this goal, new fuels and cladding materials must be developed and tested to high burnup levels (e.g. >20%) requiring cladding to withstand very high doses (greater than 200 dpa) while in contact with the coolant and the fuel. To develop and qualify materials to a total fluence greater than 200 dpa requires development of advanced alloys and irradiations in fast reactors to test these alloys. Recent results from testing numerous ferritic/martensitic steels at low temperatures suggest that improvements in low temperature radiation tolerance can be achieved through carefully controlling the nitrogen content in these alloys. Thus, four new heats of HT-9 were produced with controlled nitrogen content: two by Metalwerks and two by Sophisticated Alloys. Initial results on these new alloys are presented including microstructural analysis and hardness testing. Future testing will include irradiation testing with ions and in reactor.

36 MATERIALS SCIENCE↗
DOE OSTI2017

EPRI/DOE High Burnup Fuel Sister Pin Test Plan Simplification and Visualization

The EPRI/DOE High Burnup Confirmatory Data Project (herein called the "Demo") is a multi-year, multi-entity confirmation demonstration test with the purpose of providing quantitative and qualitative data to show how high-burnup fuel ages in dry storage over a ten-year period. The Demo involves obtaining 32 assemblies of high-burnup PWR fuel of four common cladding alloys from the North Anna Nuclear Power Plant, drying them according to standard plant procedures, and then storing them in an NRC-licensed TN-3 2B cask on the North Anna dry storage pad for ten years. After the ten-year storage time, the cask will be opened and the rods will be examined for signs of aging. Twenty-five rods from assemblies of similar claddings, in-reactor placement, and burnup histories (herein called "sister rods") have been shipped from the North Anna Nuclear Power Plant and are currently being nondestructively tested at Oak Ridge National Laboratory. After the non-destructive testing has been completed for each of the twenty-five rods, destructive analysis will be performed at ORNL, PNNL, and ANL to obtain mechanical data. Opinions gathered from the expert interviews, ORNL and PNNL Sister Rod Test Plans, and numerous meetings has resulted in the Simplified Test Plan described in this document. Some of the opinions and discussions leading to the simplified test plan are included here. Detailed descriptions and background are in the ORNL and PNNL plans in the appendices . After the testing described in this simplified test plan h as been completed , the community will review all the collected data and determine if additional testing is needed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Irradiation of Wrought FeCrAl Tubes in the High Flux Isotope Reactor

The Advanced Fuels Campaign within the Nuclear Technology Research and Development program of the Department of Energy Office of Nuclear Energy is seeking to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are one of the leading candidate materials for fuel cladding to replace traditional zirconium alloys because of the superior oxidation resistance of FeCrAl. However, there are still some unresolved questions regarding irradiation effects on the microstructure and mechanical properties of FeCrAl at end-of-life dose levels. In particular, there are concerns related to irradiation-induced embrittlement of FeCrAl alloys due to secondary phase formation. To address this issue, Oak Ridge National Laboratory has developed a new experimental design to irradiate shortened cladding tube specimens with representative 17×17 array pressurized water reactor diameter and thickness in the High Flux Isotope Reactor (HFIR) under relevant temperatures (300–350°C). Post-irradiation examination will include studies of dimensional change, microstructural changes, and mechanical performance. This report briefly summarizes the capsule design concept and the irradiation test matrix for six rabbit capsules. Each rabbit contains two FeCrAl alloy tube specimens. The specimens include Generation I and Generation II FeCrAl alloys with varying processing conditions, Cr concentrations, and minor alloying elements. The rabbits were successfully assembled, welded, evaluated, and delivered to the HFIR along with a complete quality assurance fabrication package. Pictures of the rabbit assembly process and detailed dimensional inspection of select specimens are included in this report. The rabbits were inserted into HFIR starting in cycle 472 (May 2017).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

UFD Proof-of-Concept Demonstration

The BISON Spent Nuclear Fuel modeling capability has been demonstrated through simulations of the complete nuclear fuel lifecycle including in-reactor irradiation, spent fuel pool cooling, vacuum drying and finally dry cask storage. The importance of hydrogen in the cladding and the formation of hydrides at various stages of the fuel lifecycle have been a focus of the initial development work. In addition, the integration in BISON of a hydride damage model has improved the prediction of the cladding mechanical behavior. The work described in this report was performed under funding from the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. This report is issued in satisfaction of the level 2 milestone M2MS-17IN0201012 on UFD Proof-of-Concept Demonstration for the INL Engineering-Scale Fuel Performance (BISON) Workpackage.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Establish Fuel Qualification Expectations

The objective of this white paper is to provide an overview of the state of knowledge regarding issues associated with fuel qualification for a liquid-fueled molten salt reactor. The overview consists of fuel qualification expectations for heterogeneous reactors—conventional light water reactors, plus advanced reactors such as sodium fast reactors and modular high-temperature gas-cooled reactors—which consider the mechanical behavior of the fuel itself, interaction between fuel and its cladding, behavior of the cladding, and structural behavior of fuel assemblies. Fuel qualification expectations for molten salt reactors are viewed in contrast to the fuel qualification for heterogeneous reactors in that the fuel is liquid (molten) and there is no accumulation of mechanical effects such as stress, creep, or swelling. However, the liquid fuel salts do accumulate fission products, corrosion byproducts, and buildup of actinides which impact the chemistry of the fuel salt and thermal-physical properties that may impact the ability to remove heat from the core and retain radioactive materials in the fuel salt. Finally, the role of the US Department of Energy in research on tri-structural-isotropic and metal fuel for advanced reactors for fuel qualification purposes is briefly discussed as a potential model for the Department’s research to gather quality chemical and thermal-physical property data necessary to qualify MSR fuel performance under normal and accident conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Cheesecloth Alternatives Testing

Heat source plutonium oxide (HS-PuO 2 ) enriched with 65% or greater plutonium-238, is recovered and purified using an aqueous nitric acid process. Historically, cheesecloth has been employed for small spill clean-up and for routine housekeeping. However, the 2016 Department of Energy (DOE) Operating Experience report, entitled Nitrate Waste Evaluations, stated that the mixing of nitrate wastes with organic absorbents, such as cheesecloth, may render transuranic waste unacceptable for disposal at WIPP in the future. This report explores the feasibility of replacement of cheesecloth with an alternative sorbent wipe that has similar absorption and cleaning abilities and improved acid resistance. Currently, cheesecloth is used for three different functions in HS-PuO 2 operations: spill cleanup; general housekeeping operations such as cleaning glovebox floors and windows, balances, furnaces, etc.; and decontamination of iridium cladded heat sources containing HS-PuO 2 . Alternatives used for spill cleanup must be able to tolerate high neutron doses found in the HS-Pu gloveboxes. They must also be able to withstand exposure to nitric acid concentrations up to 15.8 M without decomposing. Wipes used for general housekeeping must also be durable enough to clean rough surfaces and remain intact when used for scrubbing. Wipes used for clad decontamination do not need to absorb large amounts of liquid but they must be able to withstand neutron exposures, temperatures of up to several hundred degrees Celsius, and exposure to acid solutions followed by mild scrubbing. Finally wipes used for any purpose must resist chemical reactions with nitric acid that could cause issues for waste disposal such as energetic exothermic reactions or flammable gas build-up. The wipes selected for this study were ranked based on their usability (radiation tolerance, absorbency, integrity in nitric acid, and durability) as well as their chemical compatibility with nitric acid as demonstrated using several types of chemical analyses. Wipes were then ranked based on the combination of their usability and chemical compatibility scores.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗