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All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

Modeling of Some Physical Properties of Zirconium Alloys for Nuclear Applications in Support of UFD Campaign

Zirconium-based alloys Zircaloy-2 and Zircaloy-4 are widely used in the nuclear industry as cladding materials for light water reactor (LWR) fuels. These materials display a very good combination of properties such as low neutron absorption, creep behavior, stress-corrosion cracking resistance, reduced hydrogen uptake, corrosion and/or oxidation, especially in the case of Zircaloy-4. However, over the last couple of years, in the post-Fukushima Daiichi world, energetic efforts have been undertaken to improve fuel clad oxidation resistance during off-normal temperature excursions. Efforts have also been made to improve upon the already achieved levels of mechanical behavior and reduce hydrogen uptake. In order to facilitate the development of such novel materials, it is very important to achieve not only engineering control, but also a scientific understanding of the underlying material degradation mechanisms, both in working conditions and in storage of used nuclear fuel. This report strives to contribute to these efforts by constructing the thermodynamic models of both alloys; constructing of the respective phase diagrams, and oxidation mechanisms. A special emphasis was placed upon the role of zirconium suboxides in hydrogen uptake reduction and the atomic mechanisms of oxidation. To that end, computational thermodynamics calculations were conducted concurrently with first-principles atomistic modeling.

42 ENGINEERING↗
DOE OSTI2014

Nuclear Fuels & Materials Spotlight Volume 4

As the nation's nuclear energy laboratory, Idaho National Laboratory brings together talented people and specialized nuclear research capability to accomplish our mission. This edition of the Nuclear Fuels and Materials Division Spotlight provides an overview of some of our recent accomplishments in research and capability development. These accomplishments include: • The first identification of silver and palladium migrating through the SiC layer in TRISO fuel • A description of irradiation assisted stress corrosion testing capabilities that support commercial light water reactor life extension • Results of high-temperature safety testing on coated particle fuels irradiated in the ATR • New methods for testing the integrity of irradiated plate-type reactor fuel • Description of a 'Smart Fuel' concept that wirelessly provides real time information about changes in nuclear fuel properties and operating conditions • Development and testing of ultrasonic transducers and real-time flux sensors for use inside reactor cores, and • An example of a capsule irradiation test. Throughout Spotlight, you'll find examples of productive partnerships with academia, industry, and government agencies that deliver high-impact outcomes. The work conducted at Idaho National Laboratory helps to spur innovation in nuclear energy applications that drive economic growth and energy security. We appreciate your interest in our work here at INL, and hope that you find this issue informative.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2014

Technical Report on the Status of Nonlinear Ultrasonic Techniques for Non-Destructive Inspection

This report summarizes technical work conducted by LANL staff and international collaborators in support of the UFD Storage Experimentation effort. The overall focus of this technical work is two-fold: a) on the detection and imaging of a failure mechanism known as stress corrosion cracking (SCC) in stainless steel, and b) on evaluating concrete degradation using nonlinear ultrasonic techniques. This year we report here on three research efforts: 1) experimental effort utilizing TREND and SPACE inspection techniques for crack detection in stainless steel; 2) numerical modeling of 3D time reversal focusing to quantify TREND inspection regions; and 3) application of nonlinear ultrasonic techniques for quantifying concrete degradation. The first and third were collaborative efforts. The first was performed in conjunction with the Tohoku University (Sendai, Japan) and CNRS Parametric Imaging Laboratory (Laboratoire d’Imagerie Parametrique, in Paris, France), with support from the Electricity of France (Électricité de France, [EDF], in Paris, France) project. The third research effort was performed in conjunction with the Federal Institute for Material Research and Testing (Bundesanstalt für Materialforschung und –prüfung, in Berlin, Germany) and The Pennsylvania State University (State College, PA). The conclusion to be made from these results is that nonlinear ultrasound is applicable as a spot inspection technique in both stainless steel and concrete, with the potential to be more broadly applied to other materials and environments. The nonlinear signature measured is related to the degree of damage/degradation. The localized nature of the measurement allows for identification of mechanical defects and localization of degraded regions. Further work is necessary to characterize the degradation (e.g., sensitivity, size of flaws, type of damage, residual strength, etc.). Additionally, for the SCC work to continue to progress, relevant samples must be acquired for study.

36 MATERIALS SCIENCE↗
DOE OSTI2014

The Effect of Cold Work on Properties of Alloy 617

Alloy 617 is approved for non-nuclear construction in the ASME Boiler and Pressure Vessel Code Section I and Section VIII, but is not currently qualified for nuclear use in ASME Code Section III. A draft Code Case was submitted in 1992 to qualify the alloy for nuclear service but efforts were stopped before the approval process was completed.1 Renewed interest in high temperature nuclear reactors has resulted in a new effort to qualify Alloy 617 for use in nuclear pressure vessels. The mechanical and physical properties of Alloy 617 were extensively characterized for the VHTR programs in the 1980’s and incorporated into the 1992 draft Code Case. Recently, the properties of modern heats of the alloy that incorporate an additional processing step, electro-slag re-melting, have been characterized both to confirm that the properties of contemporary material are consistent with those in the historical record and to increase the available database. A number of potential issues that were identified as requiring further consideration prior to the withdrawal of the 1992 Code Case are also being re-examined in the current R&D program. Code Cases are again being developed to allow use of Alloy 617 for nuclear design within the rules of the ASME Boiler and Pressure Vessel Code. In general the Code defines two temperature ranges for nuclear design with austenitic and nickel based alloys. Below 427°C (800°F) time dependent behavior is not considered, while above this temperature creep and creep-fatigue are considered to be the dominant life-limiting deformation modes. There is a corresponding differentiation in the treatment of the potential for effects associated with cold work. Below 427°C the principal issue is the relationship between the level of cold work and the propensity for stress corrosion cracking and above that temperature the primary concern is the impact of cold work on creep-rupture behavior.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Application of nonlinear ultrasonics to inspection of stainless steel for dry storage

This report summarized technical work conducted by LANL staff an international collaborators in support of the UFD Storage Experimentation effort. The focus of the current technical work is on the detection and imaging of a failure mechanism known as stress corrosion cracking (SCC) in stainless steel using the nonlinear ultrasonic technique known as TREND. One of the difficulties faced in previous work is in finding samples that contain realistically sized SCC. This year such samples were obtained from EPRI. Reported here are measurements made on these samples. One of the key findings is the ability to detect subsurface changes to the direction in which a crack is penetrating into the sample. This result follows from last year's report that demonstrated the ability of TREND techniques to image features below the sample surface. A new collaboration was established with AGH University of Science and Technology, Krakow, Poland.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Development of a robust modeling tool for radiation-induced segregation in austenitic stainless steels

Irradiation-assisted stress corrosion cracking (IASCC) of austenitic stainless steels in Light Water Reactor (LWR) components has been linked to changes in grain boundary composition due to irradiation induced segregation (RIS). This work developed a robust RIS modeling tool to account for thermodynamics and kinetics of the atom and defect transportation under combined thermal and radiation conditions. The diffusion flux equations were based on the Perks model formulated through the linear theory of the thermodynamics of irreversible processes. Both cross and non-cross phenomenological diffusion coefficients in the flux equations were considered and correlated to tracer diffusion coefficients through Manning’s relation. The preferential atom-vacancy coupling was described by the mobility model, whereas the preferential atom-interstitial coupling was described by the interstitial binding model. The composition dependence of the thermodynamic factor was modeled using the CALPHAD approach. Detailed analysis on the diffusion fluxes near and at grain boundaries of irradiated austenitic stainless steels suggested the dominant diffusion mechanism for chromium and iron is via vacancy, while that for nickel can swing from the vacancy to the interstitial dominant mechanism. The diffusion flux in the vicinity of a grain boundary was found to be greatly influenced by the composition gradient formed from the transient state, leading to the oscillatory behavior of alloy compositions in this region. This work confirms that both vacancy and interstitial diffusion, and segregation itself, have important roles in determining the microchemistry of Fe, Cr, and Ni at irradiated grain boundaries in austenitic stainless steels.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Evaluation of the Frequencies for Canister Inspections for SCC

This report fulfills the M3 milestone M3FT-15SN0802042, “Evaluate the Frequencies for Canister Inspections for SCC” under Work Package FT-15SN080204, “ST Field Demonstration Support – SNL”. It reviews the current state of knowledge on the potential for stress corrosion cracking (SCC) of dry storage canisters and evaluates the implications of this state of knowledge on the establishment of an SCC inspection frequency. Models for the prediction of SCC by the Japanese Central Research Institute of Electric Power Industry (CRIEPI), the United States (U.S.) Electric Power Research Institute (EPRI), and Sandia National Laboratories (SNL) are summarized, and their limitations discussed.

25 ENERGY STORAGE↗
DOE OSTI2016

Thermodynamic and Kinetic Modeling on Thermal Segregation of Phosphorus in Iron

Radiation induced segregation (RIS) has been frequently reported in structural materials such as austenitic, ferritic, and ferritic-martensitic stainless steels (SS) that have been widely used in light water reactors (LWRs). RIS has been linked to secondary degradation effects in SS including irradiation induced stress corrosion cracking (IASCC). Earlier studies on thermal segregation in Fe based alloys found that metalloids elements such as P, S, Si, Ge, Sn etc. embrittle the materials when enrichment was observed at grain boundaries. RIS of Fe-Cr-Ni based austenitic steels has been modeled in the U.S. 2015 fiscal year (FY2015), which identified the pre-enrichment due to thermal segregation can have an important role on the subsequent RIS. The goal of this work is to develop thermal segregation models for alloying elements in steels for future integration with RIS modeling.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2016

Characterization and Modeling of Grain Boundary Chemistry Evolution in Ferritic Steels under Irradiation

Ferritic/martensitic (FM) steels such as HT-9, T-91 and NF12 with chromium concentrations in the range of 9-12 at.% Cr and high Cr ferritic steels (oxide dispersion strengthened steels with 12-18% Cr) are receiving increasing attention for advanced nuclear applications, e.g. cladding and duct materials for sodium fast reactors, pressure vessels in Generation IV reactors and first wall structures in fusion reactors, thanks to their advantages over austenitic alloys. Predicting the behavior of these alloys under radiation is an essential step towards the use of these alloys. Several radiation-induced phenomena need to be taken into account, including phase separation, solute clustering, and radiation-induced segregation or depletion (RIS) to point defect sinks. RIS at grain boundaries has raised significant interest because of its role in irradiation assisted stress corrosion cracking (IASCC) and corrosion of structural materials. Numerous observations of RIS have been reported on austenitic stainless steels where it is generally found that Cr depletes at grain boundaries, consistently with Cr atoms being oversized in the fcc Fe matrix. While FM and ferritic steels are also subject to RIS at grain boundaries, unlike austenitic steels, the behavior of Cr is less clear with significant scatter and no clear dependency on irradiation condition or alloy type. In addition to the lack of conclusive experimental evidence regarding RIS in F-M alloys, there have been relatively few efforts at modeling RIS behavior in these alloys. The need for predictability of materials behavior and mitigation routes for IASCC requires elucidating the origin of the variable Cr behavior. A systematic detailed high-resolution structural and chemical characterization approach was applied to ion-implanted and neutron-irradiated model Fe-Cr alloys containing from 3 to 18 at.% Cr. Atom probe tomography analyses of the microstructures revealed slight Cr clustering and segregation to dislocations and grain boundaries in the ion-irradiated alloys. More significant segregation was observed in the neutron irradiated alloys. For the more concentrated alloys, irradiation did not affect existing Cr segregation to grain boundaries and segregation to dislocation loops was not observed perhaps due to a change in the dislocation loop structure with increasing Cr concentration. Precipitation of α’ was observed in the neutron irradiated alloys containing over 9 at.% Cr. However ion irradiation appears to suppress the precipitation process. Initial low dose ion irradiation experiments strongly suggest a cascade recoil effect. The systematic analysis of grain boundary orientation on Cr segregation was significantly challenged by carbon contamination during ion irradiation or by existing carbon and therefore carbide formation at grain boundaries (sensitization). The combination of the proposed systematic experimental approach with atomistic modeling of diffusion processes directly addresses the programmatic need for developing and benchmarking predictive models for material degradation taking into account atomistic kinetics parameters

36 MATERIALS SCIENCE↗
DOE OSTI2016

Post-deformation examination of specimens subjected to SCC testing

This report details the results of post-radiation and post-deformation characterizations performed during FY 2015–FY 2016 on a subset of specimens that had previously been irradiated at high displacement per atom (dpa) damage doses. The specimens, made of commercial austenitic stainless steels and alloys, were subjected to stress-corrosion cracking tests (constant extension rate testing and crack growth testing) at the University of Michigan under conditions typical of nuclear power plants. After testing, the specimens were returned to Oak Ridge National Laboratory (ORNL) for further analysis and evaluation.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Fracture toughness evaluation of select advanced replacement alloys for LWR core internals

Life extension of the existing nuclear reactors imposes irradiation of high fluences to structural materials, resulting in significant challenges to the traditional reactor materials such as type 304 and 316 stainless steels. Advanced alloys with superior radiation resistance will increase safety margins, design flexibility, and economics for not only the life extension of the existing fleet but also new builds with advanced reactor designs. The Electric Power Research Institute (EPRI) teamed up with Department of Energy (DOE) to initiate the Advanced Radiation Resistant Materials (ARRM) program, aiming to develop and test degradation resistant alloys from current commercial alloy specifications by 2021 to a new advanced alloy with superior degradation resistance in light water reactor (LWR)-relevant environments by 2024. Fracture toughness is one of the key engineering properties required for core internal materials. Together with other properties, which are being examined such as high-temperature steam oxidation resistance, radiation hardening, and irradiation-assisted stress corrosion cracking resistance, the alloys will be down-selected for neutron irradiation study and comprehensive post-irradiation examinations. According to the candidate alloys selected under the ARRM program, ductile fracture toughness of eight alloys was evaluated at room temperature and the LWR-relevant temperatures. The tested alloys include two ferritic alloys (Grade 92 and an oxide-dispersion-strengthened alloy 14YWT), two austenitic stainless steels (316L and 310), four Ni-base superalloys (718A, 725, 690, and X750). Alloy 316L and X750 are included as reference alloys for low- and high-strength alloys, respectively. Compact tension specimens in 0.25T and 0.2T were machined from the alloys in the T-L and R-L orientations according to the product forms of the alloys. This report summarizes the final results of the specimens tested and analyzed per ASTM Standard E1820. Unlike the ferritic alloys showing slight decreases (Grade 92) or significant decreases (14YWT) in fracture toughness at elevated temperatures, the fracture toughness of the austenitic stainless steels and Ni-base superalloys were not strongly dependent upon the test temperatures. The fracture toughness of the alloys at the LWR-relevant temperatures was estimated by averaging the toughness values within 250– 350°C, which suggested the fracture toughness of the alloys in a descending order as 316L (752±98 MPa√m), 310 (513±66 MPa√m), 718A (313±43 MPa√m), 690 (267±48 MPa√m), 725 (218±55 MPa√m), X750 (145±16 MPa√m), Grade 92 (112±12 MPa√m), and 14YWT (63±3 MPa√m). Tearing modulus of the alloys was analyzed in the meantime, which were not strongly dependent upon the test temperatures. The high-strength alloys 718A, 725, X750, and 14YWT had the lowest tearing modulus, ranging from ~45 to ~7. Alloy 690 exhibited the highest tearing modulus on the order of 450, followed by 316L and 310 on the order of 260. Grade 92 had a noticeably lower tearing modulus on the order of 70.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Risk Assessment of Structural Integrity of Transportation Casks after Extended Storage

This study assessed the risk of loss of structural integrity of transportation casks and fuel cladding after extended storage. Although it is known that fuel rods discharged from NPPs have a small percentage of rod cladding defects, the behavior of fuel cladding and the structural elements of assemblies during transportation after long-term storage is not well understood. If the fuel degrades during extended storage, it could be susceptible to damage from vibration and impact loads during transport operations, releasing fission-product gases into the canister or the cask interior (NWTRB 2010). Degradation of cladding may occur due to mechanisms associated with hydrogen embrittlement, delayed hydride cracking, low temperature creep, and stress corrosion cracking (SCC) that may affect fuel cladding and canister components after extended storage of hundreds of years. Over extended periods at low temperatures, these mechanisms affect the ductility, strength, and fracture toughness of the fuel cladding, which becomes brittle. For transportation purposes, the fuel may be transferred from storage to shipping casks, or dual-purpose casks may be used for storage and transportation. Currently, most of the transportation casks will be the former case. A risk assessment evaluation is conducted based on results from experimental tests and simulations with advanced numerical models. A novel contribution of this study is the evaluation of the combined effect of component aging and vibration/impact loads in transportation scenarios. The expected levels of deterioration will be obtained from previous and current studies on the effect of aging on fuel and cask components. The emphasis of the study is placed on the structural integrity of fuel cladding and canisters.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

SCC Aerosol Transport Model Summary Report

The purpose of the present work is to introduce a generic and reliable numerical model for prediction of aerosol transport, deposition, and plugging in leak paths similar to stress corrosion cracks while accounting for potential plugging formation. The model is dynamic (changing leak path geometry due to plugging) and it relies on the numerical solution of the aerosol transport equation in one dimension using finite differences. An extensive validation exercise (particle diameters: 0.01-10 μm and pressure difference up to 12 kPa) was conducted based on comparisons with experimental and theoretical data. The developed model is fairly general since it is based on a generic mechanistic description of the aerosol flow and particle deposition in the leak path, without requiring experimental fittings other than using the leak rate parameters commonly available in practical applications. Despite the approximate agreement of the model with the experimental data, the obtained predictions can be qualified as realistic. This is a major achievement of the model, which has been based on generic aerosol mechanics with no reference or adjustment to experimental data.

42 ENGINEERING↗
DOE OSTI2019

Modeling and Validation of Irradiation Damage in Ni-based Alloys for Long-Term LWR Applications (Final Report)

As light water reactor (LWR) plant lives are extended, the need for predictive modeling tools for materials degradation increase in order to ensure safe operation and plan for component replacements. The Grizzly code, which is built on the MOOSE multi-physics simulation environment, is being developed for exactly this purpose. To extend Grizzly to include capabilities for modeling Ni-based alloys, we propose a US/UK integrated program to address thermal and irradiation-induced transformations mechanisms of Alloys 690 and 625. Alloy 690 is widely used in existing LWR plants due to its superior stress corrosion cracking (SCC) resistance compared to Alloy 600. Alloy 625 is used in more limited applications but offers the benefits of both high strength (in the aged condition) and corrosion/SCC resistance. Research has shown that both alloys can undergo phase changes due to thermal or irradiation exposure. In the precipitation-hardened condition, Alloy 625 “softens” during neutron irradiation as the strengthening precipitates decompose and metastable precipitates form. However, the nature and rates of these transformations as a function of exposure conditions are not well understood. Similarly, the effects of these thermal and irradiation-induced microstructural changes on mechanical properties require evaluation. The proposed program combines thermal and irradiation experiments, mechanical testing, microstructural characterization using state-of-the art analytical techniques, atomistic modeling, micro-and macro-scale modeling via Grizzly. This multi-pronged approach yields several benefits: 1) atomistic scale models capture the physics of the mechanisms of phase transformations, which can be generalized to other systems, 2) the modeling results (from atomistic to continuum) will be validated experimentally to ensure the predictive aspect, 3) the microstructure/property relationships captured with mechanical testing provides information at the engineering scale that is needed to validate the integration of the proposed new capability into Grizzly

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Coupling between TOUGH3 and FLAC3D

Sandia National Laboratories has hired Itasca Consulting Group, Inc., the authors of the FLAC3D geomechanics software, to couple FLAC3D with TOUGH3, the porous media flow solver. The work is being done to enable a coupled mechanical-thermal-hydraulic analysis of a potential criticality event in a dual purpose cannister (DPC). The U.S. Department of Energy Office of Spent Fuel and Waste Science & Technology is investigating the performance of DPCs for direct geological disposal of spent nuclear fuel. Post closure criticality control is an important aspect of this investigation. Over geological timescales, it is envisioned that the canister and canister overpack will develop fractures due to stress corrosion processes. A breach in the canister could allow groundwater to fill the canister. Fresh water is a neutron moderator; thus, if the canister internals and fuel assemblies have been sufficiently degraded, a criticality event could occur. Such an event would release enough energy to boil the water between the fuel rods and pressurize the cannister. This internal pressurization may cause the initial fractures in the canister and overpack to grow. It is important to understand the change in hydraulic transmissivity between the canister and surroundings for two reasons: first, because it may control the potential for and frequency of subsequent criticality events; second, because it will control the release of radionuclides from the canister. The motivation for this work is to better understand the potential for periodic criticality events, cannister damage, and release of radionuclides during a criticality event in a DPC.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Preliminary Test Design and Plan for a Canister Deposition Field Demonstration

This report provides a high-level test plan for deploying three commercial 32PTH2 spent nuclear fuel (SNF) canisters inside NUHOMS Advanced Horizontal Storage Modules (AHSM) from Orano (formerly Transnuclear Inc.). The details contained in this report represent the best designs and approaches explored for testing as of this publication. Given the rapidly developing nature of this test program, some of these plans may change to accommodate new objectives or adapt in response to conflicting requirements. The goal of the testing is to collect highly defensible and detailed surface deposition measurements from the surface of dry storage systems in a marine coastal environment to guide chloride-induced stress corrosion crack (CISCC) research. To facilitate surface sampling, the otherwise highly prototypic dry storage systems will not contain SNF but rather will be electrically heated to mimic the thermal-hydraulic environment. Instrumentation throughout the canister, storage module, and environment will provide an extensive amount of information for the use of model validation. Manual sampling over a comprehensive portion of the canister surface at regular time intervals will offer a high-fidelity quantification of the conditions experienced in a harsh yet realistic environment.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Measurement of Particulate Retention in Microchannel Flows

The purpose of this study was to explore the flow rates and aerosol retention of an engineered microchannel with characteristic dimensions similar to those of stress corrosion cracks (SCCs) that could form in dry cask storage systems (DCSS) for spent nuclear fuel. Additionally, pressure differentials covering the upper limit of commercially available DCSS were studied. Given the scope and resources available, these data sets should be considered preliminary and are intended to demonstrate a new capability to characterize SCC under well-controlled boundary conditions. The gap of the microchannel tested was 28.9 gm (0.00110 in.), the width was 12.7 mm (0.500 in.), and the length was 8.86 mm (0.349 in.). Over a nine-hour period, the average mass concentration upstream of the microchannel was 0.048 mg/m 3 while the average concentration downstream was 0.030 mg/m 3 . By the end of the test, the mass of aerosols that entered the test section upstream of the microchannel was 0.207 mg and the mass of aerosols that exited the microchannel was 0.117 mg. Therefore, 44% of the aerosols available for transmission was retained upstream of microchannel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Status Update for the Canister Deposition Field Demonstration

This report updates the high-level test plan for evaluating surface deposition on three commercial 32PTH2 spent nuclear fuel (SNF) canisters inside NUTECH Horizontal Modular Storage (NUHOMS) Advanced Horizontal Storage Modules (AHSM) from Orano (formerly Transnuclear Inc.) and provides a description of the surface characterization activities that have been conducted to date. The details contained in this report represent the best designs and approaches explored for testing as of this publication. Given the rapidly developing nature of this test program, some of these plans may change to accommodate new objectives or requirements. The goal of the testing is to collect highly defensible and detailed surface deposition measurements from the surface of dry storage canisters in a marine coastal environment to guide chloride-induced stress corrosion crack (CISCC) research. To facilitate surface sampling, the otherwise highly prototypic dry storage systems will not contain SNF but rather will be electrically heated to mimic the thermal-hydraulic-environment. Instrumentation throughout the canister, storage module, and environment will provide an extensive amount of information for the use of model validation. Manual sampling over a comprehensive portion of the canister surface at regular time intervals will offer a high-fidelity quantification of the conditions experienced in a harsh yet realistic environment.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗