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At least 343 records · Page 19

Transporting Non-Compliant TRU Wastes in the OPTIMUS{sup TM} Packages - 20493

The OPTIMUS{sup TM} product line of packages was developed with the purpose of providing a versatile and modular packaging option for shipping problematic wastes and fissile material contents. The product line includes a high-activity design in the OPTIMUS-H and a low activity design in the OPTIMUS-L. The two variants utilize a singular containment vessel design, that is enclosed in different protective components that offer different levels of shielding along with thermal and impact protection. The thick shielding provided by the OPTIMUS-H packaging allows for higher activity contents than the OPTIMUS-L to be transported. But the lightweight design of the OPTIMUS-L packaging allows for more packages to be transported in a single shipment. The containment vessel design shared between the two packages provides leak-tight containment of all radioactive contents along with the capabilities to easily leak test and the option to inert and backfill the contents, as necessary, prior to each shipment of the package. Together the OPTIMUS package designs offer significant flexibility and are capable of handling a wide variety of waste materials and other radioactive contents. One of the primary contents covered in the initial design effort for these packages is TRU waste contents beyond the standard waste materials that are compliant with the WIPP Waste Acceptance Criteria. More specifically, the initial content of interest is TRU Waste drums containing sealed containers with potentially flammable gases. While there are multiple other packaging options available for transporting standard WIPP compliant TRU wastes, the goal of the OPTIMUS packages is to offer a superior option for transporting both standard TRU wastes as well as the more problematic non-compliant wastes, among other contents. The non-compliant TRU waste contents explicitly included for the OPTIMUS packages are standard aerosol cans and DOT 3E lecture bottles. However, the methods applied for the package containment can be easily adjusted to cover other sealed container types with potentially flammable gases. The primary challenges with including these items are both the obvious issue of potential for flammable gases (e.g. aerosol propellants) in the package, but also the uncertainty in the exact state of the contents. Because these desired contents are waste materials, the characterization of the materials present in the waste may include some uncertainties. For example, though it may be known that there are one or more aerosol cans in a TRU waste drum, it may not be known if this can is full, spent, or anywhere in between. Also there may be equal uncertainty in the contents of the aerosol can, specifically in the potentially flammable propellant remaining in the can. To include these types of non-compliant items as acceptable contents for the package, the methods utilized in the safety analyses of the OPTIMUS packages must consider all of the uncertainties in the characterization of the waste. The methods utilized cover a range of concerns with transporting contents of this nature including pressure buildup, gas generation, and gaseous combustion for demonstrating the containment of the package. The ability to transport these problematic contents in the OPTIMUS packages gives TRU waste generating sites the option to relocate these drums to an offsite location where the non-compliant items can be properly managed through removal or destruction. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Neutron-Gamma flux and Dose calculations in a Pressurized Water Reactor (PWR)

The Institute of Radiological Protection and Nuclear Safety (IRSN) acts as technical support to the French nuclear safety authorities. In this context, one of its missions is to improve nuclear safety through the enhancement of knowledge regarding operating nuclear power reactors. The issues concerning the aging and the characterisation of the reactor vessel require knowledge of the neutron and gamma radiation environment on the vessel and beyond. Since few years, ENEA (Energia Nucleare e Energie Alternative, i.e. Nuclear Energy and Alternative Energy agency) and IRSN collaborate on Monte-Carlo simulations and Variance Reduction Techniques (VRT), which are used to determine neutron and gamma spectra in and out of the core. Additionally, IRSN participates in the DISCOMS project, which stands for 'Distributed Sensing for Corium Monitoring and Safety', aiming at examining the status of the third barrier of confinement, in order to define possible mitigation strategies in case of a severe accident, namely: i) reactor pressure vessel breakthrough and corium relocation outside the vessel, ii) reactor pit concrete erosion and iii) corium cooling. One of the IRSN tasks in the frame of the DISCOMS project is to provide the radiation level in terms of neutron and gamma doses and flux energy spectra at various locations. This paper summarizes the study performed on the radiation transport in the Tihange-I Belgian nuclear reactor and contributes to several issues requiring the knowledge of the radiation environment in and outside the reactor vessel. This paper gives an overview of the work presented at the ICRS-13 and RPSD-2016 conference (for more details see the proceedings paper)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Out-of-Pile Testing and Instrumentation Transient Water Irradiation System

Current initiatives to increase the burnup of conventional nuclear fuels past the approximate 62 GWd/t limit have been spurred on by direct savings to refueling and waste storage. The technical justification for a new license limit requires extensive qualification through experimental testing. Unlike beginning-of-life fuels, high-burnup fuels are more susceptible to fuel fragmentation, relocation, and dispersal (FFRD), therefore more data is needed to characterize fuels under key accident scenarios. The Transient Reactor Test Facility (TREAT) located at the Idaho National Laboratory has developed a testing apparatus architecture to test fuels and claddings at prototypic conditions. The Transient Water Irradiation System (TWIST) is the latest iteration of a testing device capable of conducting loss of coolant accidents (LOCAs) in TREAT. The Out-of-Pile Testing and Instrumentation TWIST (OPTI-TWIST) is an electrically heated device that is analogous to TWIST. OPTI-TWIST allows for detailed instrumentation and thermal-hydraulic characterization. TWIST ultimately aims to conduct the most advanced in-situ diagnostics to evaluate FFRD in a prototypic LOCA. Moreover, it will explore the phenomenological bifurcation of a decay-energy heat up driven LOCA and a stored-energy heat up driven LOCA. The instrumentation suite includes conventional thermocouples and pressure transducers in addition to an electro impedance sensor, an acoustic emission sensor, an optical pressure sensor, and an optical pyrometer. Characterizing these instruments in OPTI-TWIST eliminates complications of irradiation effects while preserving extreme thermal-hydraulic conditions. Finally, benchmarking both devices to a thermal-hydraulic code like the Reactor Excursion and Leak Analysis Program (RELAP)5-3D provides a unique opportunity for iteration. Pre-test predictions and post-test interpretations inform the physical designs, operational procedures, test conditions, and instrumentation types and positions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Design of a Loss of Coolant Blowdown Capsule for Remote Assembly with High Burnup Fuel

The Transient Water Irradiation System is an enhanced capability capsule type irradiation vehicle designed to support fuel safety research for light water reactor specimens in the Transient Reactor Test Facility and is designed to simulate loss of coolant and reactivity-initiated accidents. The capsule was designed, deployed, and commissioned with fresh fuel specimens to validate instrumentation and prepare for previously irradiated high burnup specimens. The irradiation system features an extensive in-situ instrumentation package to detect phenomena typical to light water reactor fuels. To accommodate the assembly with high burnup specimens inside the Hot Fuel Examination Facility, and to ensure instrumentation integrity is maintained throughout assembly, the design was updated to support remote handling. The updated design features a hinge mechanism which allows for remote pre-irradiated specimen loading and assembly while protecting sensitive instrumentation by relocating during loading. Fixtures and equipment have been developed to handle the experiment capsule and components in cell, and to support remote assembly. The experiment module also includes radiation shielding and contamination control to support operations after removal from the hot cell.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Loss of Coolant High Burnup (LOC-HBu) Program

LOC-HBu is a joint experimental program (JEEP) operating within the Nuclear Energy Agency’s (NEA’s) framework for irradiation experiments II (FIDES-II). LOC-HBu is dedicated to the understanding of light water reactor (LWR) fuel performance at high burnup under loss of coolant accident (LOCA) conditions. LOC-HBu aims to support burnup extension needs by addressing identified R&D priorities to achieve an improved understanding of fuel fragmentation, relocation, and dispersal (FFRD) of HBu fuel during LOCA events. The need for such experiments was identified in the United States by the Fuel Performance Testing Technical Expert Group (FPTTEG) under the Electric Power Research Institute (EPRI)’s Collaborative Research on Advanced Fuel Technologies (CRAFT) framework. A combined in-pile and furnace testing program plan was developed by researchers at Idaho National Laboratory (INL) and Oak Ridge National Laboratory (ORNL) which was reviewed and approved by the FPTTEG Combined TREAT-LOC and SATS LOCA Experiment Plan. The first 4 tests identified in this plan are being put forward by the U.S. Department of Energy (U.S. DOE) and INL for inclusion in the FIDES-II second triannual work period. The data produced under this plan will be used to further validate and confirm existing models and inform future R&D and model development. The experimental program was specifically developed to address data gaps and opportunities identified via detailed review of the existing public knowledge base on LOCA FFRD, as well as reviewing specific experimental development activities regarding prototypic LOCA conditions for light-water reactor (LWR) systems. The program will systematically investigate the impacts of prototypic HBu fuel/cladding thermomechanical behaviors under postulated LWR LOCA conditions not yet fully investigated. These conditions correspond with prototypic decay-energy heatup (DEH) and stored-energy heatup (SEH) conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

The Transient Reactor Test Facility (TREAT)

TREAT operated from 1959-1994, later refurbished & resumed operation in 2017 to support fuel safety testing Zircaloy-clad graphite/fuel blocks comprise core Virtually any power history possible within ~2000 MJ core transient energy capacity From milliseconds to minutes: Pulses, Ramps, LOCA Fuel motion monitoring system “hodoscope” observes fast neutrons emitted from specimens to track fuel relocation in real time Reactor also can be a neutron source to adjacent radiography facility Experiment vehicle does everything else Safety containment, specimen environment, and instrumentation

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Computational Investigation of Fuel Dispersal Phenomena during Large-Break Loss of Coolant Accident in Light-Water Reactors

In the event of cladding rupture, which could occur in light water reactor fuel assemblies during a loss-of-coolant accident (LOCA), fuel particles, along with fission gases, can be expelled into the reactor core from the fractured fuel rod. This expulsion of fragmented fuel particles, referred as fuel dispersal, is the subject of investigation to evaluate the safety implications of increasing fuel burnup in light water reactors, with a specific focus on fuel fragmentation, relocation, and dispersal. Particle trajectories and the resulting mass distribution of the settled particles within the reactor pressure vessel can pose a long-term cooling challenge for the reactor core. The fuel dispersal phenomenon is significantly influenced by the ejection characteristics of the fuel fragments, as well as the size and shape of the cladding rupture and fuel rod depressurization history during LOCA transients. In this study, the transport of fuel particles within a scaled 5 × 5 lattice of a pressurized water reactor rod bundle geometry is modeled through a two-fluid Eulerian framework that treats the gas and solid phases as interpenetrating continua. The required boundary conditions are evaluated from the fuel performance code BISON in a postulated large-break LOCA scenario. The modeling framework considers solid fuel particles as granular matter, interacting with the gaseous dry steam phase and fission gases through the governing interfacial momentum and energy exchange between the gas and solid phases. The simulation results provide the volume fraction of the solids settled on the bottom surface of the fuel bundle, quantifying the deposition within the bundle geometry.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Meso-scale modeling of UO2 nuclear fuel to high burnup

To improve the economics of light water reactors for commercial nuclear energy generation, utility operators are seeking to obtain regulatory approval to run UO2 fuel to higher levels of burnup. One potential impediment to obtaining this approval is the phenomenon of fuel fragmentation, relocation, and dispersal (FFRD). FFRD can result when fuel experiences a rapid temperature transient, such as that occurring during a Loss Of Coolant Accident (LOCA). FFRD has historically been most associated with the rim region in UO2 fuel pellets, where the phenomenon of fragmentation is also referred to as pulverization due to the small size of the fragments. More recent evidence suggests that the so-called “dark zone” (due to its appearance in micrographs) that can be observed in the mid-radial regions of high burnup fuel is also susceptible to FFRD. Although empirical fuel performance models have been developed that can adequately predict pulverization in the rim region under typical LWR conditions, a scientific understanding of what underlies fuel restructuring and subsequent FFRD is lacking even in the rim region, and no models are currently available for the behavior the dark zone. To address these challenges, the U.S. Department of Energy’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) program has employed a multi-scale modeling approach to improve scientific understanding and develop new fuel performance models. In this talk, I will focus on meso-scale efforts, which form a crucial link between atomic-scale and engineering-scale models. Phase-field modeling combined with cluster dynamics is used to predict the restructuring process in the rim region. Phase-field fracture modeling, informed by atomistic simulations, is used to predict the onset of pulverization in the rim region. Combining these techniques together allows the extent of rim pulverization to be predicted. The formation and evolution of the dark zone has also been simulated with the phase-field method, using an improved approach to vacancy source term parameterization. The work shows the important impact of microstructure on fuel performance.

fracture↗

Non-destructive Post-Transient Examination Results of the MOXTOP-1 Experiment

The MOXTOP-1 experiment has been assembled and irradiated in INL’s HFEF and TREAT facilities followed by disassembly of the experiment to enable initial post-transient examination. Results of neutron radiography of the fuel pin still in the THOR capsule were somewhat inconclusive due to the additional scatter and attenuation from the iron heat sink included in the design. Nonetheless, no obvious failure or fuel relocation could be identified in the radiography. Following disassembly of the fuel pin from the capsule, initial visual inspections were performed, and documented via images obtained through the hot cell window. No obvious signs of fuel failure were identified, indicating that the fuel pin remained intact during the transient. Some discoloration was observed near the middle of the fuel pin, likely corresponding to the fueled zone indicating that that region may have reached a relatively high temperature. Pre-transient, and post-transient profilometry of the fuel pin were performed and compared. Some additional strain displacement in the bottom portion of the fuel zone is identified, but remains well below 1%.

Mixed Oxide↗

Operationally induced preferred basis in unitary quantum mechanics

The preferred-basis problem and the definite-outcome aspect of the measurement problem persist even if the detector is modeled unitarily, because experimental data are necessarily represented in a Boolean event algebra of mutually exclusive records whereas the theoretical description is naturally formulated in a noncommutative operator algebra with continuous unitary symmetry. This change of mathematical type constitutes the core of the 'cut': a structurally necessary interface from group-based kinematics to set-based counting. In the presented view the basis relevant for recorded outcomes is not determined by the system Hamiltonian alone; it is induced by the measurement mapping, i.e., by the detector channel together with the coarse-grained readout that defines an instrument. The probabilistic mapping is anchored in symmetry and measure theory: by Gleason-type uniqueness (Gleason for projections in $d>2$ and Busch's extension for Positive Operator-Valued Measures (POVMs) including $d=2$), the trace rule is the unique probability measure consistent with additivity over exclusive events and basis-independence of the unitary sector. A compact qubit--pointer model yields an induced unsharp POVM $E_\pm=\tfrac12(\id\pm η\,σ_z)$ with $η$ fixed by pointer resolution, displaying explicitly how the detector induces the relevant basis. Finally, nested-observer paradoxes are tightened into a non-composability lemma: joint assignment of outcome propositions is obstructed unless a joint instrument exists. This relocates the origin of randomness to the stochasticity of the transition rules.

Pronskikh, Vitaly [Fermilab] (ORCID:00000002518174↗

The Fuel Motion Monitoring System at TREAT - Current Status and Future Plans

An important component of the United States Nuclear Fuel Safety Transient Testing Program, the Fuel Motion Monitoring System (FMMS) at Idaho National Laboratory's Transient Reactor Test Facility (TREAT) is fast-neutron hodoscope capable of imaging the location, movement, and relocation of nuclear fuel experiments under simulated transient accident conditions. The FMMS was refurbished in parallel with the TREAT restart program starting in 2014, restoring 96-channels of fast-neutron detection. Since returning to operation in 2017 the FMMS has supported many fuel safety experiments supporting accident tolerant fuel development, light-water reactor safety, space thermal nuclear propulsion fuel development, and advanced reactor research and development. In Phase 2 of the FMMS restoration. work is now under way to expand the FMMS' field-of-view by adding an additional 96 channels of fast neutron detectors to the system's hodoscope, along with an expanded data acquisition system and associated transient timing electronics. An overview of the FMSS system and its fast-neutron detectors will be presented along with examples of current FMMS imaging performance and associated information.

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗

Fuel Bonding and its Impact on Axial Gas Communication Behavior in Light-Water Reactor Fuel Rods

Axial gas communication concerns the flow along the axial axis of nuclear fuel rods during ramp and loss of coolant accident (LOCA) conditions. During power ramps, the higher linear heat generation rate may cause fuel-to-clad gap closure that may prevent transport of released fission gases to the plenum. Upon reduction in power the gas then can communicate to the plenum. This phenomenon has been experimentally observed by short power dips during ramp experiments completed at the Risø reactor. At higher burnups it is observed that the UO2 fuel and Zircaloy cladding forms a chemical bond. This bond results in complete closure of the gap. When these high burnup rods are subjected to a LOCA, the bond has implications on both the mechanical response (i.e., ballooning) of the cladding and subsequent fuel relocation and axial gas communication. In the LOCA scenario, gas communication is of interest in two different regimes: 1) pre-rupture communication from the plenum towards the lower pressure ballooning area and 2) the post-rupture depressurization of the plenum to the external system pressure. In both regimes the presence of a fuel-to-cladding bond will impact the rate of depressurization. In this work we present a fuel-to-clad bonding model that is coupled to an existing axial gas communication model framework in the BISON fuel performance code. The effect of considering the bond on fuel performance modeling predictions is presented through comparisons to existing experimental data. Experiments considered include several rods from the Halden IFA-650 test series. An evaluation on a full-length rod that explores the combined effect of plenum size and bonding status on axial gas communication behavior is also presented.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Technical Assistance for the Makah Tribe, Washington

The Energy Technology Innovation Partnership Project (ETIPP) worked with the Makah Tribe on strategic energy planning, including scoping future energy needs of relocated facilities, understanding the potential for new energy generation options to support current and future energy demand, and supporting communication and capacity-building goals.

24 POWER TRANSMISSION AND DISTRIBUTION↗

Waif Gopher Tortoise Survival and Site Fidelity Following Translocation

Gopher tortoises (Gopherus polyphemus) are among the most commonly translocated reptiles. While there are risks associated with any translocation, waif tortoises— animals frequently of unknown origin that have been displaced from the wild and often held in human possession for various reasons and durations—are generally excluded from translocation projects due to heightened concerns of introducing pathogens and uncertainty about the post-release survival of these individuals. If these risks could be managed, waif tortoises could have conservation value, as they can provide the needed numbers to stabilize populations. In the early 1990s, the discovery of an isolated population of gopher tortoises (≤15 individuals) near Aiken, South Carolina, USA, prioritized establishment of the Aiken Gopher Tortoise Heritage Preserve (AGTHP). Due to the population’s need for augmentation and the site’s isolation from other tortoise populations, the AGTHP provided the opportunity to evaluate the post-release survival of translocated waif tortoises without compromising a viable population. Since 2006, >260 waif tortoises have been introduced to the preserve. Using a Cormack-Jolly-Seber modeling framework to analyze release records and capture histories from trapping efforts in 2017 and 2018, we estimated the long-term apparent survival and site fidelity of this population composed largely of waif tortoises. We estimated annual apparent survival probabilities to be high (≥0.90) for subadult, adult male, and adult female tortoises, and these rates were similar to those reported for wild-to-wild translocated gopher tortoises and those from unmanipulated populations. Of the tortoises recaptured within the boundaries of the preserve, 75% were located within 400 m of their release location. Furthermore, these results suggest that waif tortoises could be an important resource in reducing the extirpation risk of isolated populations.

60 APPLIED LIFE SCIENCES↗

A Eulerian three-fluid flow framework for the study of fuel dispersal behavior under loss-of-coolant accident conditions

As the nuclear industry explores new fuel designs to accommodate increased burnup, studying fuel behavior during loss-of-coolant accidents is essential for ensuring the safe operation of light-water reactors. Here, this study presents a computational model to simulate the intricate three-phase flow dynamics of fuel dispersal following cladding breach, where solid fuel particles and carrier fission gases interact with the surrounding liquid or gas medium. The model utilizes a Eulerian three-fluid framework to depict the phase interaction within interpenetrating continua, treating fragmented fuel as granular material. Closure relationships for bulk-flow properties of the solid phase and detailed interfacial exchange terms in high-temperature, high-speed gas-solid-liquid flow scenarios are proposed. The model was validated using adiabatic experimental results in the literature, focusing on characterizing fuel dispersal behavior during loss-of-coolant accidents conditions. Settlement characteristics of the solids, quantified by the probability distribution of equivalent particles, closely matched probability density functions reported in experimental studies. Key highlights of this study include the theoretical description of a coupled three-phase flow with phase change, interpretation of the high-pressure boundary conditions and insights into the transient behavior of gas-solid-liquid phase dynamics during a simulated high-speed dispersal event.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Computational model development and validation of fuel dispersal phenomena

As the US nuclear industry is renewing efforts to extend the rod average burnup limits in existing pressurized water reactors, it becomes imperative to investigate the safety implications associated with high burnup fuel fragmentation and the potential dispersal of fragmented fuel into the reactor coolant system. Additionally, as there is a growing interest in transitioning to new nuclear fuel designs to accommodate higher fuel burnup levels, it becomes essential to conduct studies on fuel behavior during design basis accident scenarios, particularly during a postulated loss-of-coolant accident, in order to ensure the safe operation of light-water reactors. The present study develops a computational model to simulate the complex three-phase flow of the fuel dispersal phenomena that occur following a breach in the cladding, when the stream of solid fuel particles and carrier fission gas stream may interact with the surrounding liquid medium. Furthermore, the proposed framework considers the solid phase as discrete Lagrangian particles while the interpenetrating gas–liquid continuum is modeled using a Eulerian framework. The developed simulation framework is validated with experimental results performed in a separate-effect test facility by comparing predicted particle settlement locations with the experimental observations reported in the literature. Key highlights of this study include the interpretation of the high-pressure boundary conditions and the three-phase flow coupling strategy. Also discussed are the gas–liquid hydrodynamics and the behavior of particles as they are transported through the gas and liquid phases.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Washington State Play Fairway Analysis - Passive Monitoring of St. Helens Shear Zone for Tomography and Precision Microseismic Event Detection

Data resources were derived from a passive seismic survey of the northern St. Helens Shear Zone on geothermal leases 12-24 km north of Mount St. Helens for phase 2 of the Geothermal Play-Fairway Analysis of Washington State Prospects. A 20 seismic station array of broadband seismometers was deployed with irregular spacing (1-4 km) over an area of 12 km to image seismogenic features and their damage zones in the shallow crust.

15 GEOTHERMAL ENERGY↗