Brady Hanson

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Papers by Brady Hanson

Developing the technical basis for long term storage and subsequent transportation of used nuclear fuel

OSTI OAI (U.S. Department of Energy Office of Scientific and Technical Information), Dec 1, 2011

of GLOBAL 2011 Nagoya, Japan, Sep 4-9, 2011 Developing the Technical Basis for Long Term Storage ... more of GLOBAL 2011 Nagoya, Japan, Sep 4-9, 2011 Developing the Technical Basis for Long Term Storage and Subsequent Transportation of Used Nuclear Fuel Ken Sorenson, Brady Hanson, Felicia Duran, Paul McConnell, and Ruth Weiner Sandia National Laboratories, Pacific Northwest National Laboratory ABSTRACT: The United States Department of Energy (DOE), Office of Nuclear Energy (NE) is engaged in a program to develop the technical bases for extending dry storage and subsequent transportation of used nuclear fuel. With the cancellation of the Yucca Mountain repository project, the DOE/NE is currently evaluating alternative fuel cycles, including alternative concepts to dispose of wastes produced by candidate fuel cycles. Since this evaluation is planned to take decades, the need to store these fuels past their regulatory certification periods has become clear. And, since development of the technical basis to extend certification storage periods will take multiple years, the time to start addressing the technical needs is now. The NE program addressing this issue is divided into four main topical areas; R&D Opportunities, Security, Transportation, and Concept Evaluations. R&D Opportunities is focused on conducting a data gap analysis that will identify the technical issues that need to be addressed in order to demonstrate fuel integrity in a storage environment over long periods of time. Security will identify and address security issues that arise as a consequence of long term storage that may not be apparent over a much shorter period of time. Transportation is focused on technical issues associated with transport of fuel after extended storage, as well as transport of high burnup fuel in the near term. Concept Evaluations will roll up these three efforts and develop a process for evaluation of candidate alternatives that will be considered for gathering the needed data. As part of the NE program, close collaboration with industry and international organizations is essential. This collaboration includes participation on the Electric Power Research Institute (EPRI) Extended Storage Collaboration Program (ESCP) Committee. This committee is charted to identify technical gaps and associated means to gather the necessary data. In addition, ESCP has a subcommittee that is focused on international participation in order to get the broadest input from across the world on issues concerning long term storage and transportation of used fuel. It is expected that significant leverage can be attained by combining efforts on similar issues from organization across the world. This paper will provide an overview of the NE program, a status of the work performed to date, and a vision of future work The United States Department of Energy (DOE), Office of Nuclear Energy (NE) is engaged in a program to develop the technical bases for extending dry storage and subsequent transportation of used nuclear fuel. With the cancellation of the Yucca Mountain repository project, the DOE/NE is currently evaluating alternative fuel cycles, including alternative concepts to dispose of wastes produced by candidate fuel cycles. Since this evaluation is planned to take decades, the need to store these fuels past their regulatory certification periods has become clear. And, since development of the technical basis to extend certification storage periods will take multiple years, the time to start addressing the technical needs is now. The NE program addressing this issue is divided into four main topical areas; R&D Opportunities, Security, Transportation, and Concept Evaluations. R&D Opportunities is focused on conducting a data gap analysis that will identify the technical issues that need to be addressed in order to demonstrate fuel integrity in a storage environment over long periods of time. Security will identify and address security issues that arise as a consequence of long term storage that may not be apparent over a much shorter period of time. Transportation is focused on technical issues associated with transport of fuel after extended storage, as well as transport of high burnup fuel in the near term. Concept Evaluations will roll up these three efforts and develop a process for evaluation of candidate alternatives that will be considered for gathering the needed data. As part of the NE program, close collaboration with industry and international organizations is essential. This collaboration includes participation on the Electric Power Research Institute (EPRI) Extended Storage Collaboration Program (ESCP) Committee. This committee is charted to identify technical gaps and associated means to gather the necessary data. In addition, ESCP has a subcommittee that is focused on international participation in order to get the broadest input from across the world on issues concerning long term storage and transportation of used fuel. It is expected that significant leverage can be attained by combining efforts on similar issues from organization…

Neptunium(V) Partitioning to Uranium(VI) Oxide and Peroxide Solids

Environmental Science & Technology, May 4, 2005

Metaschoepite, [(UO2)8O2(OH)12] x 10H2O, and metastudtite, UO4 x 4H2O, are alteration phases anti... more Metaschoepite, [(UO2)8O2(OH)12] x 10H2O, and metastudtite, UO4 x 4H2O, are alteration phases anticipated in a spent nuclear fuel repository following the moist oxidation of UO2 on a geologic time scale. Dissolved concentrations and hence potential mobility of other radionuclides in the fuel, such as the neptunyl cation (NpO2+), will likely be determined by the extent of their partitioning into these U(VI) solids. 237Np is of particular interest due to its potential high mobility and long half-life (2.1 x 10(6) years.) In this study, metaschoepite has been precipitated and subsequently transformed to studtite in the presence of dissolved Np. The metaschoepite and studtite solids that formed initially contained &amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;lt;10 and 6500 ppm Np, respectively. Batch dissolution studies of these solids at pH 6 demonstrate release of Np that exceeds congruent dissolution of U from metastudtite; furthermore, the released Np cation remains in solution. Thus, although the Np partitions into the metastudtite solid initially, it is released to solution over time, indicating that metastudtite is not likely to serve as a host solid for Np incorporation or sorption of the neptunyl cation on long time scales.

Presented at the ASTM 24th Symposium on Effects of Radiation on Nuclear Materials and the Nuclear Fuel Cycle

ASTM International eBooks, 2010

Thermal and Deposition Modeling of the Canister Deposition Field Demonstration

A Review of NDE Methods for Detecting and Monitoring of Atmospheric SCC in Dry Cask Storage Canisters for Used Nuclear Fuel

Corrosion, Mar 17, 2013

Dry cask storage systems (DCSSs) for used nuclear fuel (UNF) were originally envisioned for stora... more Dry cask storage systems (DCSSs) for used nuclear fuel (UNF) were originally envisioned for storage periods of short duration (~ a few decades). However, uncertainty challenges the opening of a permanent repository for UNF implying that UNF will need to remain in dry storage for much longer durations than originally envisioned (possibly for centuries). Thus, aging degradation of DCSSs becomes an issue that may not have been sufficiently considered in the design phase and that can challenge the efficacy of very long-term storage of UNF. A particular aging degradation concern is atmospheric stress corrosion cracking (SCC) of DCSSs located in marine environments. In this report, several nondestructive (NDE) methods are evaluated with respect to their potential for effective monitoring of atmospheric SCC in welded canisters of DCSSs. Several of the methods are selected for evaluation based on their usage for in-service inspection applications in the nuclear power industry. The technologies considered include bulk ultrasonic techniques, acoustic emission, visual techniques, eddy current, and guided ultrasonic waves.

Fuel assembly shaker and truck test simulation

This report fulfills the requirement of milestone M3FT-14PN0813034 "Shaker Table Report" under wo... more This report fulfills the requirement of milestone M3FT-14PN0813034 "Shaker Table Report" under work package FT-14PN081303. This study continues the modeling support of the Sandia National Laboratories (SNL) shaker table task from 2013 and includes analysis of the SNL 2014 truck test campaign. Detailed finite element models of the fuel assembly surrogate used by SNL during testing form the basis of the modeling effort. Additional analysis was performed to characterize and filter the accelerometer data collected during the SNL testing. The detailed fuel assembly finite element model was modified to improve the performance and accuracy of the original surrogate fuel assembly model in an attempt to achieve a closer agreement with the low strains measured during testing. The revised model was used to recalculate the shaker table load response from the 2013 test campaign. As it happened, the results remained comparable to the values calculated with the original fuel assembly model. From this it is concluded that the original model was suitable for the task and the improvements to the model were not able to bring the calculated strain values down to the extremely low level recorded during testing. The model needs more precision to calculate strains that are so close to zero.

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Evaluation of Radiolytic Hydrogen Generation in the High Burnup Demonstration Dry Storage Cask

OSTI OAI (U.S. Department of Energy Office of Scientific and Technical Information), Feb 17, 2021

Residual water (post-drying and sealing) in a spent nuclear fuel (SNF) dry storage canister/cask ... more Residual water (post-drying and sealing) in a spent nuclear fuel (SNF) dry storage canister/cask could include: • free water trapped as a liquid and/or water as a vapor; • physisorbed/chemisorbed water that is bound to internal surfaces, such as the cask and internals including fuel rods and aluminum components. This water is bound with varying adsorption energy, typically with the first few monolayers strongly bound with weakly bound layers forming on top in equilibrium with the humidity; and • chemisorbed water that is chemically bound in an (oxy)hydroxide film. These waters are subject to radiolysis from the attendant radiation of the fuel. The free and physisorbed waters radiolytically break down into molecular hydrogen and hydrogen peroxide, an oxidizing species, that itself would decompose into hydrogen and oxygen, with the oxygen expected to react with zirconium fuel cladding, resulting in net generation of hydrogen gas and ZrO2. Radiolytic breakdown of chemicallybound water in (oxy)hydroxides can also occur to generate hydrogen gas. The sources and amounts of residual water and the radiolysis thereof to cause hydrogen generation has been estimated for the High Burnup Demonstration (HBU Demo) Cask, a TN-32 cask loaded with high-burnup Light Water Reactor spent fuel assemblies and vacuum-dried at the North Anna Power Station in November 2017. The cask internals include 32 assemblies, each with 264 fuel rods clad in zirconium alloys, aluminum neutron absorber components, and aluminum and stainless-steel structural components. Empirical information from the literature for the radiolytic hydrogen yield in terms of G-values a has been applied to estimate the net H2 expected from the separate radiolytic generation of the residual water inventories. Backreactions in the helium cover gas are not expected, and gettering of H2 was assumed to be negligible due to material condition of a hydride rim (hydrogen-saturated condition) on the HBU fuel cladding surface. The results for the estimation in terms of H2 concentration in the cask volume in ppmv at two points in time are shown in the table below. For the 12-day time, two cases were considered: 1) case 1 assumes that the physisorbed/chemisorbed water present on the cladding surface following dry-out is not replenished by water vapor in the cask; 2) case 2 assumes that water vapor in the cask quickly replenishes the cladding surface water due to high affinity to the ZrO2 on its surface. The radiolytic hydrogen generated at 12 days duration was estimated at 534 ppmv for Case 1, in good agreement to the ~500 ppmv measured from a gas sample from the HBU cask. This paper describes the bases for residual water and hydrogen estimation in the HBU Demo cask. The application of the results shows there are no canister pressurization or flammability issues with extended dry storage of SNF. a The G-value is typically given in units of molecules of H2 per 100 eV or in units of µmol of H2 per Joule for energy deposited into the water-bearing material component. 1 μmol/joule = 9.63 molecules/100 eV WM2021 Conference,

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Technical basis for long-term storage of used nuclear fuel

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Data Analysis of Plutonium Sorption on Colloids in a Minimal Kinetics Model

Yucca Mountain is the U.S. Department of Energy's potential geologic repository designed to store... more Yucca Mountain is the U.S. Department of Energy's potential geologic repository designed to store and dispose of commercial spent nuclear fuel, defense high level waste glass, and other high-level radioactive waste forms. If approved, the site would be the nation's first geological repository for disposal of this type of radioactive waste. Yucca Mountain is located on federal land adjacent to the Nevada Test Site in Nye County, Nevada. It is approximately 100 miles northwest of Las Vegas, Nevada. Colloids may have the potential to transport strongly sorbing radionuclide contaminants in soils and groundwater aquifers . Recent studies from the Nevada Test Site, a site with a similar geology to Yucca Mountain, have indicated the enhanced mobility of plutonium in the saturated zone, albeit in minute quantities, in association with various silicate minerals (Kersting et al. 1999). The observed association of plutonium with silicate colloids by Kersting et al. (1999), sorption experiments by Lu et al. (2000), and waste form corrosion tests ) indicate that the attachment of plutonium to colloids may be an irreversible process. However, current Yucca Mountain models that assume irreversible sorption of plutonium, americium, and thorium onto colloids have led to highly conservative scenarios. This is partly because of the K d ranges used as well as the total surface area assumed to be available for uptake. Nevertheless, a slowly reversible term may provide more realistic predictions for radionuclide transport in the proposed repository. This report considers a sorption model containing the minimal dynamic features of the system to fit plutonium adsorption data similar to that developed by . Global fits to recent data favored nonzero values of reversible sorption, allowing the definition of equilibrium distribution coefficients in all cases except the synthetic form of montmorillonite. In most cases, the two-site model was adequate to fit the data. The model represents a mathematic simplification of the time-dependent sorption process and takes no account for pH-dependent surface charge changes and actinide-mineral surface interfacial chemistry. However, this allows the model to be readily incorporated into existing performance assessment codes when applied to repository environment relevant data sets.

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Grain Size and Burnup Dependence of Spent Fuel Oxidation: Geological Repository Impact

MRS Proceedings, 1999

Further refinements to the oxidation model of Stout et al. have been made. The present model inco... more Further refinements to the oxidation model of Stout et al. have been made. The present model incorporates the burnup dependence of the oxidation rate in addition to an allowance for a distribution of grain sizes. The model was tested by comparing the model results with the oxidation histories of spent fuel samples oxidized in Thermogravimetric Analysis(TGA) or Oven Dry-Bath (ODB) experiments. The comparison between the experimental and model results are remarkably close and confirm the assumption that grain-size distributions and activation energies are the important parameters to predicting oxidation behavior. The burnup dependence of the activation energy was shown to have a greater effect than decreasing the effective grain size in suppressing the rate of the reaction U 4 O 9 →U 3 O 8 . Model results predict that U 3 O 8 formation of spent fuels exposed to oxygen will be suppressed even for high burnup fuels that have undergone restructuring in the rim region, provided the repository temperature is kept sufficiently low.

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Cladding Degradation Model

An Integrated Approach to Closing the Technical Data Gap for High Burnup Spent Fuel Performance during Normal Conditions of Transport (NCT)

Initial Results From Dissolution Testing of Spent Fuel Under Acidic Conditions

MRS Proceedings, 2004

Flowthrough dissolution tests using solutions with pH in the range 2 to 7 have been conducted on ... more Flowthrough dissolution tests using solutions with pH in the range 2 to 7 have been conducted on a moderate burnup Light Water Reactor spent fuel. Such low pH conditions have been modeled as possibly occurring in a failed waste package at the proposed repository at Yucca Mountain. The release oftotal uranium, 99Tc, 90Sr, 137Cs, and 239&amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;amp;240Pu were measured for up to 90% total reaction of the specimens. The reaction rates, determined both from the cumulative release and the release normalized to surface area, were found to decrease with increasing pH and with increasing extent of reaction. The implications to instantaneous release and long-term behavior ina geologic repository are discussed.

Dry Air Oxidation of Commercial Spent Nuclear Fuel

OSTI OAI (U.S. Department of Energy Office of Scientific and Technical Information), Feb 8, 2005

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Comparison of an Ultrasonic Phased Array Evaluation with Destructive Analysis of a Documented Leak Path in a Nozzle Removed from Service

Non-destructive and destructive testing methods were employed to evaluate a documented boric acid... more Non-destructive and destructive testing methods were employed to evaluate a documented boric acid leakage path through an Alloy 600 control rod drive mechanism (CRDM) penetration from the North Anna Unit 2 reactor pressure vessel head that was removed from service in 2002. A previous ultrasonic inservice-inspection (ISI) conducted by industry prior to the head removal, identified a probable leakage path in Nozzle 63 located in the interference fit between the penetration tube and the vessel head. In this current examination, Nozzle 63 was examined using phased array (PA) ultrasonic testing with a 5.0-MHz, eight-element annular array; immersion data were acquired from the nozzle inner diameter (ID) surface. A variety of focal laws were employed to evaluate the signal responses from the interference fit region. These responses were compared to responses obtained from a mockup specimen that was used to determine detection limits and characterization capabilities for wastage and boric acid presence in the interference fit region. Nozzle 63 was destructively examined after the completion of the ultrasonic nondestructive evaluation (NDE) to visually assess the leak paths. These destructive and nondestructive results compared favorably.

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Fuel Assembly Shaker Test Simulation

This report describes the modeling of a fuel assembly under dynamic shock loading in support of t... more This report describes the modeling of a fuel assembly under dynamic shock loading in support of the Sandia National Laboratories (SNL) shaker test campaign. The focus of the test campaign is on evaluating the response of used fuel to shock and vibration loads that can occur during highway transport. Modeling began in 2012 using an LS-DYNA fuel assembly model that was first created for modeling impact scenarios. SNL's proposed test scenario was simulated through analysis and the calculated results helped guide the instrumentation and other aspects of the testing. During fiscal year (FY) 2013, the fuel assembly model was refined to better represent the test surrogate. Analysis of the proposed loads suggested the frequency band needed to be lowered to attempt to excite the lower natural frequencies of the fuel assembly. Despite SNL's expansion of lower frequency components in their five shock realizations, pretest predictions suggested a very mild dynamic response to the test loading. After testing was completed, one specific shock case was modeled, using recorded accelerometer data to excite the model. Direct comparison of predicted strain in the cladding was made to the recorded strain gauge data. The magnitude of both sets of strain (calculated and recorded) is very low, compared to the expected yield strength of the Zircaloy-4 material. The model was accurate enough to predict that no yielding of the cladding was expected, but its precision at predicting micro strains is questionable.

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EFRT M-12 Issue Resolution: Caustic Leach Rate Constants from PEP and Laboratory-Scale Tests

The entries (other than the WTP target) show the 95% confidence interval around the median, as ca... more The entries (other than the WTP target) show the 95% confidence interval around the median, as calculated by 500 Monte Carlo realizations. Estimates of laboratory error in measurements were the basis for estimating all error. (a) The projections of boehmite leach factor were made by WTP and transmitted to

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US/EURATOM I-NERI 2019 Annual Review Assessing

Dosimetry Modeling for Predicting Radiolytic Production at the Spent Fuel-Water Interface

The radiolysis of water in contact with spent nuclear fuel (SNF) will produce oxidants and reduct... more The radiolysis of water in contact with spent nuclear fuel (SNF) will produce oxidants and reductants that can affect the dissolution of the fuel in a geologic disposal site. These products are created by initial radiolytic species which are a function of the type of radiation being emitted by the SNF, i.e. alpha, beta and/or gamma, as well as the energy of this radiation, the fuel grain size (and resulting surface-to-volume ratio) and the fuel-to-water ratio. These products interact with the surface of the fuel, creating new species and ultimately affecting the dissolution rate. The objective of the work reported here is to develop a systematic dosimetry model to determine the dose to water from the SNF as a function of these variables. This dose is calculated for different radiation types as a function of decay for the average fuel composition expected at Yucca Mountain. From these dose calculations the production rate of initial radiolytic products can be estimated. This data provides the basis for subsequent determination of the resulting chemical interactions at the fuel/water interface predicted by published theoretical and experimental data.

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Review of NDE Methods for Detection and Monitoring of Atmospheric SCC in Welded Canisters for the Storage of Used Nuclear Fuel

Dry cask storage systems (DCSSs) for used nuclear fuel were originally envisioned for storage per... more Dry cask storage systems (DCSSs) for used nuclear fuel were originally envisioned for storage periods of short duration (approximately a few decades). However, uncertainty challenges the opening of a permanent repository for used nuclear fuel, implying that used nuclear fuel will need to remain in dry storage for much longer durations than originally envisioned. Thus, aging degradation of DCSSs becomes an issue that may not have been sufficiently considered in the design phase and that can challenge the efficacy of very long-term storage of used nuclear fuel. A particular aging degradation concern is atmospheric stress corrosion cracking (SCC) of DCSSs located in marine environments. In this report, several nondestructive examination methods are evaluated with respect to their potential for effective monitoring of atmospheric SCC in welded canisters of DCSSs. The technologies considered include bulk ultrasonic techniques, acoustic emission, visual techniques, eddy current, and guided ultrasonic waves. Many of the nondestructive examination methods considered are screening technologies and are suitable for the detection of atmospheric SCC, but not for sizing of detected cracks. Eddy current techniques are considered to have good potential for detecting atmospheric SCC flaws in welded canisters, and the conditions for which eddy current techniques may be used to depth size atmospheric SCC flaws should be enumerated. Bulk ultrasonic techniques may have the best potential for depth sizing detected atmospheric SCC flaws. The only other technology considered to have potential for depth sizing of detected flaws is guided ultrasonic waves, although considerable development is required for the application to DCSS canisters and to understand the sensitivity and accuracy of the techniques. Visual techniques can be used to detect cracks, but not to determine their depth. Further, the reliability of visual techniques for detecting atmospheric SCC flaws is a potential concern. Finally, acoustic emission may be used to continuously monitor for atmospheric SCC and may be especially desirable for monitoring regions of the canister that are difficult to access. However, the sensitivity of AE to atmospheric SCC is unclear. In addition to a summary assessment for each technology, cross-cutting needs that would benefit several of the technologies assessed, and several monitoring strategies are identified. v Summary Dry cask storage systems (DCSSs) for used nuclear fuel are intended for temporary storage. Uncertainty regarding the feasibility of opening a repository for eventual disposal implies that DCSSs will need to store used nuclear fuel for periods much longer than originally envisioned. Meanwhile, the structural and leak-tight integrity of DCSSs must not be compromised by their extended period of use. Validating the integrity of DCSSs through monitoring or inspection enables management of the effects of degradation or aging that may occur over the extended storage period and which may threaten the integrity of the DCSS. The Code of Federal Regulations requires that storage confinement systems have the capability for continuous monitoring to enable maintenance of safe storage conditions. A gap analysis was recently performed under the U.S. Department of Energy's Used Fuel Disposition Campaign to support extended storage of used nuclear fuel (Hanson et al. 2012). This report highlighted monitoring of DCSSs as a high-priority gap area to ensure the safe long-term storage of used nuclear fuel. Many different monitoring needs were highlighted in the previous gap analysis and potential needs included technologies to monitor the environmental and structural integrity of DCSS systems, structures, and components, both internal and external to the confinement barrier. High priority was given to monitoring needs associated with detecting degradation of the confinement barrier. A specific concern is the potential for degradation of DCSSs located in coastal areas as a result of atmospheric stress corrosion cracking (SCC). SCC has been a persistent form of degradation in the nuclear power industry, particularly in operating nuclear power plants. Although atmospheric SCC in DCSSs and SCC in lightwater reactors is a consequence of different environmental factors, many of the nondestructive examination methods used for detection and characterization of SCC occurrences in nuclear power plants should have relevance to the detection and characterization of atmospheric SCC in DCSSs. The methods evaluated in this report include bulk ultrasonics, acoustic emission, visual techniques, eddy current, and guided ultrasonic waves. A summary of the evaluation is given later in Table 5.1. A few cross-cutting needs that are applicable to improving examinations are identified as follows:  Development of a robotic crawler or scanner is needed to encode the location of nondestructive examination data on the canister surface, maintain a consistent scan rate, maintain a consistent probeto-canister surface distance, and minimize the impact of human factors. This need is associated with bulk ultrasonic, visual, eddy current, and circumferentially guided wave methods. Robotic mechanisms are also needed for the placement of acoustic emission and axially guided wave transducers even though the sensors are not scanned over the surface. UNF used nuclear fuel UT ultrasonic testing Vg group velocity Vp phase velocity VT visual testing xi Contents

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