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At least 91 records · Page 5
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Key results from examinations of seven high burnup pressurized water reactor spent nuclear fuel rods

At present, spent nuclear fuel (SNF) assemblies discharged from US commercial power plants are placed into dry storage following a short cooling time (<10 years) in the plant’s spent fuel pool. The process of packaging the spent fuel into dry-storage canisters includes a drying step to remove residual water from the canister. During the drying process, the fuel rod cladding may reach temperatures as high as 400°C. Oak Ridge National Laboratory (ORNL) is performing destructive examinations of high burnup (HBU) (>45 GWd/MTU) SNF rods to address knowledge and data gaps related to extended interim storage and eventual transportation for disposal. The rods examined include four different kinds of fuel rod cladding: standard Zircaloy-4 (Zirc-4), low-tin (LT) Zirc-4, ZIRLO, and M5. Three rods were subjected to a thermal transient to assess the effects of decay-heat-driven high temperatures expected during vacuum drying of the fuel as it is prepared for interim dry storage. The examinations focus on the composite fuel rod performance, as compared with the performance of defueled rod cladding, and establish the baseline mechanical properties of a fuel rod before interim dry storage. The key results of these examinations are presented, including the measured mechanical and fatigue properties, observations of cladding hydrogen pickup and hydride reorientation effects on rod performance, effects of the simulated drying temperatures on rod performance, and general conclusions of SNF performance in extended interim dry storage and transport. The rods were found to be strong and durable in the expected loading conditions, even considering the formation of radial hydrides associated with vacuum drying. The combined testing provides a broad body of data supporting extended interim storage and transportation performance of HBU spent fuel.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Key Results from Detailed Nondestructive Examinations of 25 Pressurized Water Reactor High Burnup Spent Nuclear Fuel Rods

The High Burnup Spent Fuel Data Project, sponsored by the U.S. Department of Energy Office of Nuclear Energy, is focused on understanding the effects of long-term storage and transportation on high burnup (HBU) (>45 GW days per tonne uranium) light water reactor fuel. The project includes 32 HBU spent nuclear fuel (SNF) assemblies (the project assemblies) that are stored in a typical independent spent fuel storage installation (ISFSI) and 25 “sister rods”—9 SNF rods that were removed from the fuel assemblies prior to insertion to the ISFSI and 16 SNF rods removed from similar HBU assemblies. The sister rods provide a baseline of the condition of the HBU rods before loading, drying, and long-term dry storage. The project assemblies will be inspected after 10 years, and the physical state of the stored rods will be compared with the condition of the sister rods to identify any changes in physical properties during the dry storage period. Our report focuses on key results from the nondestructive postirradiation examinations of the sister rods and summarizes the results of detailed visual examinations, gamma scans, dimensional measurements, and eddy current liftoff measurements of the combined Chalk River unidentified deposits and oxide layer on the waterside surface of the rod. Additionally the data are used to calculate fuel rod and pellet stack growth rates, estimated remaining fuel rod plenum volumes, and the percentage change in fuel rod cladding diameter.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Experimental Study on Surrogate Nuclear Fuel Rods under Reversed Cyclic Bending

The mechanical behavior of spent nuclear fuel (SNF) rods under reversed cyclic bending or bending fatigue must be understood to evaluate their vibration integrity in a transportation environment. This is especially important for high-burnup fuels (>45 GWd/MTU), which have the potential for increased structural damage. It has been demonstrated that the bending fatigue of SNF rods can be effectively studied using surrogate rods. In this investigation, surrogate rods made of stainless steel (SS) 304 cladding and aluminum oxide pellets were tested under load or moment control at a variety of amplitude levels at 5 Hz using the Cyclic Integrated Reversible-Bending Fatigue Tester developed at Oak Ridge National Laboratory. The behavior of the rods was further characterized using flexural rigidity and hysteresis data, and fractography was performed on the failed rods. The proposed surrogate rods captured many of the characteristics of deformation and failure mode observed in SNF, including the linear-to-nonlinear deformation transition and large residual curvature in static tests, PPI and PCMI failure mechanisms, and large variation in the initial structural condition. Rod degradation was measured and characterized by measuring the flexural rigidity; the degradation of the rigidity depended on both the moment amplitude applied and the initial structural condition of the rods. It was also shown that a cracking initiation site can be located on the internal surface or the external surface of cladding. Finally, fatigue damage to the bending rods can be described in terms of flexural rigidity, and the fatigue life of rods can be predicted once damage model parameters are properly evaluated. The developed experimental approach, test protocol, and analysis method can be used to study the vibration integrity of SNF rods in the future.

cladding failure↗
DOE OSTI2016

ROD INTERNAL PRESSURE QUANTIFICATION AND DISTRIBUTION ANALYSIS USING FRAPCON

The discharge rod internal pressure (RIP) and cladding hoop stress (CHS) distributions are quantified forWatts BarNuclearUnit 1 (WBN1) fuel rods by modeling core cycle design data, intercycle assembly movements, operation data (including modeling significant trips and downpowers), and as-built fuel enrichments and densities of each fuel rod in FRAPCON-3.5. An alternate model for the amount of helium released from zirconium diboride (ZrB2) integral fuel burnable absorber (IFBA) layers is derived and applied to FRAPCON output data to quantify the RIP and CHS for these fuel rods. SCALE/Polaris is used to quantify fuel rod-specific spectral quantities and the amount of gaseous fission products produced in the fuel for use in FRAPCON inputs. Fuel rods with ZrB2 IFBA layers (i.e., IFBA rods) are determined to have RIP predictions that are elevated when compared to fuel rod without IFBA layers (i.e., standard rods) despite the fact that IFBA rods often have reduced fill pressures and annular fuel blankets. Cumulative distribution functions (CDFs) are prepared from the distribution of RIP predictions for all standard and IFBA rods. The provided CDFs allow for the determination of the portion of WBN1 fuel rods that exceed a specified RIP limit. Lastly, improvements to the computational methodology of FRAPCON are proposed.

cladding hoop stress↗
DOE OSTI2018

Fabrication and properties of biodegradable ZnO nano-rods/porous Zn scaffolds

Highlights: • ZnO nano-rods/porous Zn scaffolds were fabricated successfully. • Compressive property of ZnO nano-rods/porous Zn scaffolds is not decrease obviously. • The ZnO nano-rods/porous Zn scaffolds have a good corrosion resistance in SBF. • The ZnO nano-rods/porous Zn scaffolds can efficiently induce Ca-P precipitation. • The ZnO nano-rods/porous Zn scaffolds have an excellent antibacterial activity. - Abstract: Biodegradable ZnO nano-rods/porous Zn scaffolds were successfully fabricated by treating the porous Zn scaffolds in NaCl solution at 80 °C. X-ray diffraction (XRD) analysis shows that the nano-rods in situ fabricated on the porous Zn scaffolds in NaCl solution are hexagonal ZnO phase. This result is also confirmed by the image of high resolution transmission electron microscope (HRTEM) and the analysis of X-ray photoelectron spectroscopy (XPS). Compressive tests suggest that the compressive mechanical properties of the ZnO nano-rods/porous zinc scaffolds do not decrease significantly compared with that of the porous Zn scaffolds without treating in NaCl solution. Immersion tests in simulated body fluid (SBF) show that the ZnO nano-rods/porous zinc scaffolds slightly gain in weight after 16 days of immersion. Moreover, these ZnO nano-rods/porous zinc scaffolds can efficiently induce Ca-P precipitation during immersion tests and show a good bioactivity. In addition, antibacterial tests imply that the ZnO nano-rods/porous zinc scaffolds have an excellent antibacterial ability against Escherichia coli and Staphylococcus aureus. However, the results of cytotoxicity tests of ZnO nano-rods/porous Zn scaffolds are not satisfactory.

77 NANOSCIENCE AND NANOTECHNOLOGY↗
DOE OSTI2026

Design and fabrication of a four-rod radio frequency quadrupole with a variable aperture

Radiofrequency quadrupole (RFQ) linear accelerators with variable aperture cell parameters offer an effective means of accelerating high-intensity ion beams exceeding 100 mA and have been implemented in various RFQ designs. However, in a four-rod RFQ, the use of dielectric perturbation methods to measure the voltage distribution behind the rod electrodes is not applicable when implementing variable aperture geometries, because changes in the rod shape influence the electric field at the location of the perturbation element. Therefore, it was necessary to design the rod profile for each cell such that the electric field distribution behind the rods remains unaffected by changes in the aperture. In this study, we present a method to adjust the voltage distribution by modifying the geometry of the rod electrodes according to each cell's aperture parameters. This design approach enables the integration of variable aperture cells into a four-rod RFQ linac, significantly enhancing the achievable beam current. To ensure uniform capacitance and maintain a constant rod-electrode voltage despite variations in the average aperture radius, we developed a systematic method for shaping the rod electrodes. This method was refined using two-dimensional (2D) and three-dimensional (3D) electromagnetic field simulations to optimize the electrode cross-sectional profiles. The fabricated rod electrodes were then installed in a four-rod RFQ, and the electric field distribution was measured using the perturbation method and tuned by adjusting the base plates. As a result, the measured variation of ±1.1 % confirms that this design methodology can produce a sufficiently uniform voltage distribution for high-current beam acceleration. These results demonstrate the practical feasibility of a variable-aperture four-rod RFQ linac.

43 PARTICLE ACCELERATORS↗
DOE OSTI2018

Damaged Fuel Rod Service with German Turnkey Technology NCS 45 Transport Casks and Under-Water-Brazing Technology - 18182

Transportation of damaged irradiated fuel rods with the package NCS 45 developed and operated by DAHER Nuclear Technologies GmbH (DNT) saves considerable technical and financial resources in the German phase-out scenario. The technology can be used world-wide to speed up decommissioning of NPPs. In each German reactor (and most likely in all reactors in the world) a small number of damaged fuel rods have accumulated over the operation phase. In general, transport and storage solutions for fuel are concentrated on the bulk of material, i. e. complete fuel assemblies containing undamaged fuel rods. It is hence a rather straightforward process to remove almost all the fuel from the reactor pools. However, damaged fuel rods - which have to be removed from the pool prior to dismantling - can cause considerable delay in dismantling and high costs for safe-guarding and specialized solutions, even though they only contain a comparatively very tiny amount of fuel. DNT has all required tools, equipment and experience to mitigate the described problem: - The package NCS 45 licensed for the transport of damaged Uranium Oxide (UOX) and Mixed Oxide (MOX) fuel rods with a burn-up of up to 120 GWd/MgHM; - A patented underwater brazing (encapsulation) technology to render very high burn-up damaged fuel rods fit for transport; - Qualified personnel to plan, license and operate such encapsulation process and the NCS 45 in a NPP and to carry out the transports; - Proven record of successful removal of damaged fuel rods from several shut-down reactors in Germany. The package NCS 45 is currently licensed in several European countries. Due to its state-of-technique design - type B(U)F-96 approval - it could be licensed and used for world-wide applications. The underwater brazing technology was developed and patented by DNT to address the challenge of transporting high burn-up and/or damaged UOX and MOX fuel rods. Above a burn-up of 55 GWd/MgHM there are neither enough data to assess the behavior of the cladding under stress nor sufficient benchmarks available to validate the source terms of the fuel. Hence, high burn-up fuel rods are encapsulated in the fuel pool by using the underwater-brazing-technology before they are loaded into the package NCS 45. Furthermore, the capacity of the NCS 45 to transport damaged fuel rods can be considerably enhanced by encapsulating these rods. (authors)

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

Large Eddy Simulation of turbulent flow in rod bundle with spacer grids

Pressurized water nuclear reactor fuel assemblies are made of fuel rods held by spacer grids, which are the core components of the nuclear power plants. Hot fuel rods are cooled by the coolant water flow, and the spacer grids with mixing vanes distort the mainstream, producing the swirling flow which can enhance the heat transfer to improve CHF (Critical Heat Flux). Heat transfer between fuel rods and coolant is a crucial part of thermal-hydraulic analysis. For the purpose to improve the thermal hydraulic performance of the fuel assemblies, many experiments have been conducted to understand the flow structures and three-dimensional behaviors of the turbulence in the rod bundle sub-channels and then to optimize the design of the spacer grid. In the experimental work, which requires lots of efforts to get reliable data, LDA (Laser Doppler Anemometry) and PIV (Particle Image Velocimetry) are generally used to measure the velocity profile and turbulent intensity in the sub-channels. Computational Fluid Dynamics (CFD) is another useful way to study the characteristics of the detailed flow condition and heat transfer phenomena in the subchannels. The main issue of the single-phase CFD is to accurately predict the flow and turbulence characteristics. RANS method is generally used to simulate the induced lateral velocity and the heat transfer process. With the development of computer technology, some more precise but previously time-consuming turbulent models, such as LES (Large Eddy Simulation), are gradually used for the simulation of rod bundle flow with current advanced CPU resources. Ikeno and Kajishima explained the decay of swirling turbulent flow in the rod bundle using LES method. Benhamadouche acquired the turbulence induced forces on the central rod of the rod bundle through performing the LES of the flow downstream the spacer grid. Chang and Tavoularis conducted hybrid simulations with SAS SST and LES of a 5 x 5 rod bundle, and several vortices were identified in the vicinity of the vanes and the generation mechanisms, evolution progress and influence on the velocities of these vortices were clarified. In the paper we present CFD simulations of turbulent flow in rod bundle sub-channels with spacer grids of two different experiments. One is MATiS-H benchmark experiment with scaled spacer grids, which is conducted by KAERI, and is used as a CFD benchmark test, and the other is a 5 x 5 rod bundle with typical spacer grids using PIV technology, with the cooperation of Harbin Engineering University. The main objectives of the calculations are to validate LES methodology by the benchmark experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Characterization of Fuel and Cladding In and Near the Pellet-Pellet Gap of a High-Burnup Pressurized Water Reactor Fuel Rod

Oak Ridge National Laboratory (ORNL) is performing extensive destructive examinations of 15 high-burnup (HBU) spent nuclear fuel (SNF) rods from the North Anna Power Station (NAPS), which is operated by Dominion Energy Virginia [1]. The examinations are being conducted to obtain a baseline condition of the HBU rods before dry storage and are focused on understanding overall SNF rod strength and durability [2,3]. The HBU rods, referred to as sister rods or sibling rods, are similar to rods placed into dry storage at NAPS that are planned to be examined after one decade. The sister rods include several ZIRLO®-clad rods manufactured by Westinghouse Electric Company. One of the ZIRLO®-clad rods examined includes a pellet-pellet gap of 1 mm that was identified during the nondestructive examinations (NDEs) [4]. The rod was sectioned axially at that elevation to reveal the pellet-pellet interfaces and the pellet-pellet gap. The section was mounted and polished to reveal the distribution of hydride precipitates in the cladding above, below, and within the gap [5]. Total cladding hydrogen measurements will be performed to quantify the total cladding hydrogen through the gap and any additional in-solution or precipitated hydrogen in the pellet-pellet gap region

Montgomery, Rose↗
DOE OSTI2020

Sodium Void Worth and Control Rod Worth Measurements in ZPPR 15

Sodium void worth and control rod worth were measured in Phases A, B, C and D of the ZPPR15 experimental program. Sodium voiding was simulated by replacing stainless steel cans containing sodium by empty stainless steel cans of the same sizes. The control rod worth measurements were measurements of simulated control rods (CRs) and simulated empty control rod positions (CRPs). A CRP consisted of a 2 x 2 array of drawers filled with sodium plates. A CR consisted of a 2 x 2 array of drawers containing alternating columns of boron carbide plates and sodium plates in the front 18 inches and just sodium plates in the back 18 inches. The ZPPR staff published the measured sodium void worths and control rod worths and the statistical uncertainties in the measured values. In reality, the published uncertainty for each of these measurements is only one component of the total uncertainty. There are additional contributions from uncertainties related to measurement technique, plate compositions and reproducibility. However, the dominant uncertainty for the sodium void worth and control rod worth measurements is the uncertainty in βeff which is required to convert from the natural measurement units, cents, to the calculation units, pcm or Δ/k 1 k 2 . A full uncertainty analysis was performed for the sodium void worth and control rod worth measurements. The significant uncertainties were quantified, and a total uncertainty was derived for each sodium void worth or control rod worth measurement. The ZPPR-15 experimental records were used to create detailed as-built Monte Carlo models of the sodium void worth and control rod worth configurations. These models were used to compute the sodium void worths and control rod worths. (C/E – 1) and the related uncertainty in (C/E – 1) were derived for the measured worths.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2012

Structural Integrity Testing Method for PRSEUS Rod-Wrap Stringer Design

NASA Langley Research Center and The Boeing Company are developing an innovative composite structural concept, called PRSEUS, for the flat center section of a future environmentally friendly hybrid wing body (HWB) aircraft. The PRSEUS (Pultruded Rod Stitched Efficient Unitized Structure) concept uses dry textile preforms for the skins, frames, and stiffener webs. The highly loaded stiffeners are made from precured unidirectional carbon/epoxy rods and dry fiber preforms. The rods are wrapped with the dry fiber preforms and a resin infusion process is used to form the rod-wrap stiffeners. The structural integrity of the rod-wrap interface is critical for maintaining the panel s high strength and bending rigidity. No standard testing method exists for testing the strength of the rod-wrap bondline. Recently, Boeing proposed a rod push-out testing method and conducted some preliminary tests using this method. This paper details an analytical study of the rod-wrap bondline. The rod-wrap interface is modeled as a cohesive zone for studying the initiation and growth of interfacial debonding during push-out testing. Based on the correlations of analysis results and Boeing s test data, the adequacy of the rod-wrap testing method is evaluated, and potential approaches for improvement of the test method are proposed.

Wang, John T.↗
DOE OSTI2016

Preliminary Content Evaluation of the North Anna High Burn-Up Sister Fuel Rod Segments for Transportation in the 10-160B and NAC-LWT

The U.S. Department of Energy’s (DOE’s) Used Fuel Disposition Campaign (UFDC) Program has transported high-burnup nuclear sister fuel rods from a commercial nuclear power plant for purposes of evaluation and testing. The evaluation and testing of high-burnup used nuclear fuel is integral to DOE initiatives to collect information useful in determining the integrity of fuel cladding for future safe transportation of the fuel, and for determining the effects of aging, on the integrity of UNF subjected to extended storage and subsequent transportation. The UFDC Program, in collaboration with the U.S. Nuclear Regulatory Commission and the commercial nuclear industry, has obtained individual used nuclear fuel rods for testing. The rods have been received at Oak Ridge National Laboratory (ORNL) for both separate effects testing (SET) and small-scale testing (SST). To meet the research objectives, testing on multiple 6 inch fuel rod pins cut from the rods at ORNL will be performed at Pacific Northwest National Laboratory (PNNL). Up to 10 rod equivalents will be shipped. Options were evaluated for multiple shipments using the 10-160B (based on 4.5 rod equivalents) and a single shipment using the NAC-LWT. Based on the original INL/Virginia Power transfer agreement, the rods are assumed to 152 inches in length with a 0.374-inch diameter. This report provides a preliminary content evaluation for use of the 10-160B and NAC-LWT for transporting those fuel rod pins from ORNL to PNNL. This report documents the acceptability of using these packagings to transport the fuel segments from ORNL to PNNL based on the following evaluations: enrichment, A2 evaluation, Pu-239 FGE evaluation, heat load, shielding (both gamma and neutron), and content weight/structural evaluation.

10-160B↗
DOE OSTI2022

Control rod modeling in liquid metal-cooled fast reactors

Control rod modeling in Liquid Metal-cooled Fast Reactors (LMFRs) is important for an accurate simulation, especially in depletion calculations. Recently, control rod search and cusping models have been added to the LUPINE multiphysics fast reactor simulator. LUPINE stands for the 'LMFR Utility for Physics Informed Nuclear Engineering' and is currently being developed at North Carolina State University. LUPINE models the coupled multiphysics effects in LMFRs, including neutronics, thermal hydraulics, thermal expansion, and depletion. The control rod search has been implemented using a Newton-secant search in an inexact-Newton iteration and the cusping model uses a polynomial technique to correct for control rod cusping. The control rod cusping and search models were demonstrated by modeling the Advanced Burner Reactor (ABR) MET-1000 Sodium-cooled Fast Reactor (SFR) and a long-life Lead-cooled Fast Reactor (LFR) based on a Westinghouse Electric Company, LLC (WEC) design. A differential control rod worth curve was calculated for both reactor models to demonstrate the control rod cusping model. The SFR and LFR models were used to demonstrate the importance of modeling control rod movement during depletion calculations and the adverse effect of control rods on cycle length is demonstrated. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Post-irradiation Examinations of High Burnup PWR Rods - 18456

The High Burnup Spent Fuel Data Project (herein referred to as the 'project'), sponsored by the US Department of Energy (DOE) Office of Nuclear Energy (NE), is focused on experimentally evaluating the effects of long-term storage and transportation on high burnup (HBU) (defined as >45 gigawatt days per metric ton uranium) pressurized water reactor (PWR) spent nuclear fuel (SNF) [1]. To support this project, in the fall of 2017, an instrumented bolted-lid cask (project cask) at the North Anna reactor site was loaded with intact, HBU PWR fuel assemblies (project assemblies) having four different kinds of cladding and will be reopened in approximately 10-years for inspection. To support pre- and post-storage SNF rod characterization, 25 'sister rods' that are similar to the rods being stored in the project cask were extracted from seven different fuel assemblies at the North Anna nuclear power plant for post-irradiation examination (PIE). The sister rod testing program has multiple objectives, including: 1) Baseline testing to provide essential information on the physical condition and properties of the cladding and the fuel in the cladding prior to the loading, drying, and long-term dry storage process (analogous to an experiment control group). Similar tests are expected to be performed at the end of the 10-year storage period on rods extracted from the cask to identify any changes in the properties of the fuel rods that may have occurred during the dry storage period; and 2) Tests to provide material property and physical data using accelerated separate effects tests (SETs) and small-scale tests (SSTs) to explore a wide range of conditions that may occur for other dry storage systems and/or while the fuel is being transported after aging. Testing will be focused on understanding overall SNF rod strength and durability, and be comprised of both composite fuel and defueled clad testing with a focus on measuring cladding material and mechanical properties. Nondestructive testing is currently underway and plans are being updated to support the initial set of destructive tests that are expected to begin by early 2018. While the data generated can be used for multiple purposes, a primary goal of the sister rod PIE data and associated measured mechanical properties are expected to support existing fuel storage licensing and relicensing activities by addressing identified knowledge gaps, and enhance the technical bases for post-storage transportation, handling, and consolidation activities [1]. This paper presents results obtained from the sister rod testing to date, and describes some of the planned measurements, equipment, and procedures that will be implemented for the initial set of destructive examinations. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Synthesis and characterization of nano ZnO rods via microwave assisted chemical precipitation method

A microwave assisted chemical precipitation method has been employed for the synthesis of nano zinc oxide rods by reacting zinc nitrate and potassium hydroxide. The amount of potassium hydroxide was adjusted for three different pHs to achieve ZnO nano rods with varying aspect ratio. The mechanism of growth of nano rods is explained briefly. The average crystallite size of the as synthesized samples was analyzed by means of powder XRD pattern and estimated to vary from 25.6 nm to 43.1 nm. The existence of rods was confirmed using scanning electron microscopy (SEM). The samples were also analyzed using FT-IR. The optical properties of the samples were also studied by means of UV-visible spectra and Room Temperature Photo Luminescence studies. The band gap of the samples was determined from the DRS spectrum. A strong near band emission peaks due to surface defects are observed in the PL spectrum. - Graphical abstract: At the solution pH of 11 and 9, tetrapod-like and flower-like ZnO nano rods were formed along with separated rods respectively due to the formation of activated nuclei of different sizes. Highlights: Black-Right-Pointing-Pointer Increase in alkalinity of the precursor solution results in longer rods. Black-Right-Pointing-Pointer Beyond a saturation limit, the excess of added OH{sup -} ions inhibited the growth of rods. Black-Right-Pointing-Pointer Keeping all parameters the same, the alkalinity can only modify the aspect ratio of the rods and not their morphology.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Positional ordering induced by dynamic steric interactions in superparamagnetic rods

The dynamic motion produced by precessing magnetic fields can drive matter into far-from-equilibrium states. We predict 1D periodic ordering in systems of precessing rods when magnetic interactions between rods remain insignificant. The precession angle of the rods is completely determined by the field's precession angle and the ratio of the field's precession frequency and the characteristic response frequency of the rods. Here, we develop a molecular dynamics model that explicitly calculates magnetic interactions between particles, and we also simulate rods in the limit of a strong and fast precessing magnetic field where inter-rod magnetic interactions are negligible, using a purely steric model. Our simulations show how steric interactions drive the rods from a positionally disordered phase (nematic) to a layered (smectic) phase. As the rod precession angle increases, the nematic–smectic transition density significantly decreases. The minimization of unfavorable steric interactions also induces phase separation in binary mixtures of rods of different lengths. This effect is general to any force that produces precession in elongated particles. This work will advance the understanding and control of out-of-equilibrium soft matter systems.

42 ENGINEERING↗
DOE OSTI2020

Sister Rod Destructive Examinations (FY20)

As a part of DOE NE High Burnup Spent Fuel Data Project, Oak Ridge National Laboratory (ORNL) is performing destructive examinations (DEs) of high burnup (HBU) (>45 GWd/MTU) spent nuclear fuel (SNF) rods from the North Anna Nuclear Power Station operated by Dominion Energy. The SNF rods, called sister rods or sibling rods, are all HBU and include four different kinds of fuel rod cladding: standard Zircaloy-4 (Zirc-4), low-tin Zirc-4, ZIRLO, and M5. The DEs are being conducted to obtain a baseline of the HBU rod’s condition before dry storage and are focused on understanding overall SNF rod strength and durability. Composite fuel and defueled cladding will be tested to derive material properties. Although the data generated can be used for multiple purposes, one primary goal for obtaining the post-irradiation examination data and the associated measured mechanical properties is to support SNF dry storage licensing and relicensing activities by (1) addressing identified knowledge gaps and (2) enhancing the technical basis for post-storage transportation, handling, and subsequent disposition. This report documents the status of the ORNL Phase I DEs of 8 sister rods and outlines the DE tasks performed and the data collected to date, as guided by the sister rod test plans.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Sister Rod Destructive Examinations (FY22) Appendix E: Mechanical Testing - Spent Fuel and Waste Disposition

As a part of the DOE NE High Burnup Spent Fuel Data Project, Oak Ridge National Laboratory (ORNL) is performing destructive examinations (DEs) of high burnup (HBU) (>45 GWd/MTU) spent nuclear fuel (SNF) rods from the North Anna Nuclear Power Station operated by Dominion Energy. The SNF rods, called sister rods or sibling rods, are all HBU and include four different kinds of fuel rod cladding: standard Zircaloy-4 (Zirc-4), low-tin Zirc-4, ZIRLO, and M5. The DEs are being conducted to obtain a baseline of the HBU rod’s condition before dry storage and are focused on understanding overall SNF rod strength and durability. Composite fuel and defueled cladding will be tested to derive material properties. Although the data generated can be used for multiple purposes, one primary goal for obtaining the post-irradiation examination data and the associated measured mechanical properties is to support SNF dry storage licensing and relicensing activities by (1) addressing identified knowledge gaps and (2) enhancing the technical basis for post-storage transportation, handling, and subsequent disposition. This report documents the status of the ORNL Phase 1 DE activities related to the mechanical testing of selected sister rods in Phase 1 of the sister rod test program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗