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

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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Partial defect verification of spent fuel assemblies by PDET: Principle and field testing in Interim Spent fuel Storage Facility (CLAB) in Sweden

Here, the need for the development of a credible method and instrument for partial defect verification of spent fuel has been emphasized over a few decades in the safeguards communities as the diverted spent fuel pins can be the source of nuclear terrorism or devices. The need is increasingly more important and even urgent as many countries have started to transfer spent fuel to so called "difficult-to-access" areas such as dry storage casks, reprocessing or geological repositories. Partial defect verification is required by IAEA before spent fuel is placed into "difficult-to-access" areas. Earlier, Lawrence Livermore National Laboratory (LLNL) has reported the successful development of a new, credible partial defect verification method for pressurized water reactor (PWR) spent fuel assemblies without use of operator data, and further reported the validation experiments using commercial spent fuel assemblies with some missing fuel pins. The method was found to be robust as the method is relatively invariant to the characteristic variations of spent fuel assemblies such as initial fuel enrichment, cooling time, and burn-up. Since then, the PDET system has been designed and prototyped for 17×17 PWR spent fuel assemblies, complete with data acquisition software and acquisition electronics. In this paper, a summary description of the PDET development followed by results of the first successful field testing using the integrated PDET system and actual spent fuel assemblies performed in a commercial spent fuel storage site, known as Central Interim Spent fuel Storage Facility (CLAB) in Sweden will be presented. In addition to partial defect detection initial studies have determined that the tool can be used to verify the operator declared average burnup of the assembly as well as intra-assembly bunrup levels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Methodology and Software for Gross Defect Detection of Spent Nuclear Fuel at the Atucha-I Reactor [Novel Methodology and Software for Spent Fuel Gross Defect Detection at the Atucha-I Reactor]

Here, fuel assemblies in the spent fuel pool are stored by suspending them in two vertically stacked layers at the Atucha Unit 1 nuclear power plant (Atucha-I). This introduces the unique problem of verifying the presence of fuel in either layer without physically moving the fuel assemblies. Given that the facility uses both natural uranium and slightly enriched uranium at 0.85 wt% 235 U and has been in operation since 1974, a wide range of burnups and cooling times can exist in any given pool. A gross defect detection tool, the spent fuel neutron counter (SFNC), has been used at the site to verify the presence of fuel up to burnups of 8000 MWd/t. At higher discharge burnups, the existing signal processing software of the tool was found to fail due to nonlinearity of the source term with burnup.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Nanoparticle Precipitation in Irradiated and Annealed Ceria Doped with Metals for Emulation of Spent Fuels

Epsilon alloy precipitates have been observed with varied compositions and sizes in spent nuclear fuels, such as UO2. Presence of the inclusions, along with other oxide precipitates, gas bubbles and irradiation-induced structural defects, can significantly degrade the physical properties of the fuel. To predict fuel performance, a fundamental study of the precipitation processes is needed. This study uses ceria (CeO2) as a surrogate for UO2. Polycrystalline CeO2 films doped with Mo, Ru, Rh, Pd and Re (surrogate for Tc) were grown at 823 K using pulsed laser deposition, irradiated at 673 K with He+ ions, and subsequently annealed at higher temperatures. A number of methods, including transmission electron microscopy and atom probe tomography, were applied to characterize the samples. The results indicate that there is a uniform distribution of the doped metals in the as-grown CeO2 film. Pd particles of ~3 nm in size appear near dislocation edges after He+ ion irradiation to ~13 dpa. Thermal annealing at 1073 K in air leads to formation of precipitates with Mo and Pd around grain boundaries. Further annealing at 1373 K produces 70 nm sized precipitates with small grains at cavities.

Jiang, Weilin↗
DOE OSTI2011

Shipper/receiver difference verification of spent fuel by use of PDET

Spent fuel storage pools in most countries are rapidly approaching their design limits with the discharge of over 10,000 metric tons of heavy metal from global reactors. Countries like UK, France or Japan have adopted a closed fuel cycle by reprocessing spent fuel and recycling MOX fuel while many other countries opted for above ground interim dry storage for their spent fuel management strategy. Some countries like Finland and Sweden are already well on the way to setting up a conditioning plant and a deep geological repository for spent fuel. For all these situations, shipments of spent fuel are needed and the number of these shipments is expected to increase significantly. Although shipper/receiver difference (SRD) verification measurements are needed by IAEA when the recipient facility receives spent fuel, these are not being practiced to the level that IAEA has desired due to lack of a credible measurement methodology and instrument that can reliably perform these measurements to verify non-diversion of spent fuel during shipment and confirm facility operator declarations on the spent fuel. In this paper, we describe a new safeguards method and an associated instrument, Partial Defect Tester (PDET), which can detect pin diversion from Pressurized Water Reactor (PWR) Spent Fuel Assemblies in an in-situ condition. The PDET uses multiple tiny neutron and gamma detectors in the form of a cluster and a simple, yet highly precise, gravity-driven system to obtain underwater radiation measurements inside a Pressurized Water Reactor (PWR) spent fuel assembly. The method takes advantage of the PWR fuel design which contains multiple guide tubes which can be accessed from the top. The data obtained in such a manner can provide spatial distribution of neutron and gamma flux within a spent fuel assembly. Our simulation study as well as validation measurements indicated that the ratio of the gamma signal to the thermal neutron signal at each detector location normalized to the peak ratio of all the detector locations gives a unique signature that is sensitive to missing pins. The signature is principally dependent on the geometry of the detector locations, and little sensitive to enrichment or burn-up variations. A small variation in the fuel bundle, such as a few missing pins, changes the shape of the signature to enable detection. After verification of the non-diversion of spent fuel pins, the neutron signal and gamma signal are subsequently used to verify the consistency of the operator declaration on the fuel burn-up and cooling time. (authors)

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

Automated defect detection in spent nuclear fuel using combine Cerenkov and Gamma Emission Tomography data

Spent fuel monitoring and characterization has been a pillar of safeguards and nuclear facility monitoring for many years. The Digital Cerenkov Viewing Device (DCVD) has been used since the 1980s as a method of defect detection in spent fuel. In recent years, the accounting of large quantities of spent fuel before storage has renewed interest in this relatively quick and inexpensive method. This has an impact not only in safeguards, but also in for nuclear power facilities, as accounting can be a long and arduous process. Additionally, the DCVD demonstrates limited accuracy in more complex cases such as substitution of a fuel rod with steel, or a partial defect detection. A second method, Gamma Emission Tomography (GET) has been explored as an improved defect detection method, but is much more expensive and invasive than DCVD. This work identifies deficiencies in both methods, and proposes a combination of data gathered from each method to address these deficiencies for improved spent fuel characterization. Initial results are promising, showing 97% detection of a single missing fuel rod when the data types are combined, versus approximately 50% and 70%respectively for DCVD and GET data on their own. Finally, these classification results are obtained with algorithms derived from facial recognition and applied to this problem, yielding unique accuracy in near real-time while also maintaining the information barrier between output and measurement desired in safeguards.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Measurement and Analysis of Gamma-Rays Emitted From Spent Nuclear Fuel Above 3 MeV

The gamma-ray spectrum of spent nuclear fuel in the 3- to 6-MeV energy range is important for active interrogation since emitted gamma rays emitted from nuclear decay are not expected to interfere with measurements in this energy region. There is, unfortunately, a dearth of empirical measurements from spent nuclear fuel in this region. This work is an initial attempt to partially ll this gap by presenting an analysis of gamma-ray spectra collected from a set of spent nuclear fuel sources using a high-purity germanium detector array. This multi-crystal array possesses a large collection volume, providing high energy resolution up to 16 MeV. The results of these measurements establish the continuum count-rate in the energy region between 3- and 6-MeV. Also assessed is the potential for peaks from passive emissions to interfere with peak measurements resulting from active interrogation delayed emissions. As one of the first documented empirical measurements of passive emissions from spent fuel for energies above 3 MeV, this work provides a foundation for active interrogation model validation and detector development.

delayed gamma rays↗
DOE OSTI2018

A facile recovery process for cathodes from spent lithium iron phosphate batteries by using oxalic acid

A simplistic and novel leaching process is developed to dispose spent lithium iron phosphate (LiFePO 4 ) batteries. In this paper, oxalic acid is selected as a leaching reagent to recover lithium as a resource and remove phosphorus from LiFePO 4 batteries, benefiting from its low natural effects. The physical properties of spent cathode materials (before leaching) and residues (after leaching) are identified by X-ray diffraction (XRD) and scanning electronic microscopy (SEM) with energy dispersive X-ray spectroscopy (EDS). The total amounts of Li and Fe are analyzed by inductively coupled plasma-atomic emission spectroscopy (ICP-AES). The process occurs under conditions of a 0.3 mol·L -1 oxalic acid concentration, a temperature of 80°C, a reaction time of 60 min and a solid/liquid ratio of 60 g· L -1 . It can effectively precipitate ~92% ferrum in terms of FeC 2 O 4 ·2H 2 O from LiFePO 4 , with the leaching efficiency of Li achieving up to 98%. This method demonstrates a new strategy for dealing with spent lithium iron phosphate batteries with a low cost and low environmental impact.

25 ENERGY STORAGE↗
DOE OSTI2020

Fast Determination of Lithium Content in Spent Cathodes for Direct Battery Recycling

Determination of the lithium ion content in cathode materials, which is currently performed by inductively coupled plasma mass spectrometry (ICP-MS), involves tedious procedures and error-prone preparation. This work introduces a new methodology to quickly determine the lithium content in spent cathode materials (LiNix Mny Coz O2, x + y + z = 1, NMC) using a thermalgravimetric analysis, which bypasses the use of ICP-MS and significantly reduces the cost and time required for determining the lithium content. Based on this new method, a new approach is developed to identify spent cathode materials suitable for direct relithiation recycling. Finally, these new methods have been applied to investigate the effectiveness of direct electrochemical-relithiation on recycling the spent cathodes, and provide insights for future development of direct battery materials recycling.

cathode recycling↗
DOE OSTI2018

Amplicon Sequencing Reveals Microbiological Signatures in Spent Nuclear Fuel Storage Basins

Water quality is an important determinant for the structural integrity of alloy cladded fuels and assemblies during long-term wet storage. Detailed characterization of a water filled storage basin for spent nuclear reactor fuel was performed following the formation and proliferation of an amorphous white flocculent. White precipitant was sampled throughout the storage basin for chemical and spectroscopic characterization, and eDNA was extracted for pyrosequencing of bacterial rRNA gene diversity. Accordingly, spectroscopic analyses indicated the precipitant to be primarily amorphous to crystalline aluminum (oxy) hydroxides with minor associated elemental components including Fe, Si, Ti, and U. High levels of dissolved carbon were co-localized with the precipitant relative to bulk water. Bacterial densities were highly variable between sampling locations and with depth; cell numbers (log scale) ranged from 5.6 to 4.89 cells / mL. Bacterial diversity that was physically associated with the aluminum (oxy) hydroxide complexes exceeded an estimated 4,000 OTUs / amplicon library (3% cutoff) and the greatest percent majority of sequences were aligned to the families Burkholderiales (23%), Nitrospiraceae (23%), Hyphomicrobiaceae (17%), and Comamonadaceae (6%). We surmise that episodic changes in the physical and chemical properties of the basin contribute to the polymerization of aluminum (oxy) hydroxides, which in turn can chemisorb nutrients, carbon ligands and bacterial cells from the surrounding bulk aqueous phase. As such, these precipitants should establish favorable microhabitats for bacterial colonization and growth. Comparative analyses of 16S rRNA gene amplicon libraries across diverse environmental landscapes were performed and microbiological signatures unique to the spent nuclear fuel storage basin environment were revealed. These insights could spur the development of tractable bioindicators that are specific of and diagnostic for water quality at discrete locations and finer scales of resolution, marking an important contribution for improved water quality and management of spent nuclear fuel storage facilities.

aluminum↗
DOE OSTI2015

BWR Spent Nuclear Fuel Integrity Research and Development Survey for UKABWR Spent Fuel Interim Storage

The objective of this report is to identify issues and support documentation and identify and detail existing research on spent fuel dry storage; provide information to support potential R&D for the UKABWR (United Kingdom Advanced Boiling Water Reactor) Spent Fuel Interim Storage (SFIS) Pre-Construction Safety Report; and support development of answers to questions developed by the regulator. Where there are gaps or insufficient data, Oak Ridge National Laboratory (ORNL) has summarized the research planned to provide the necessary data along with the schedule for the research, if known. Spent nuclear fuel (SNF) from nuclear power plants has historically been stored on site (wet) in spent fuel pools pending ultimate disposition. Nuclear power users (countries, utilities, vendors) are developing a suite of options and set of supporting analyses that will enable future informed choices about how best to manage these materials. As part of that effort, they are beginning to lay the groundwork for implementing longer-term interim storage of the SNF and the Greater Than Class C (CTCC) waste (dry). Deploying dry storage will require a number of technical issues to be addressed. For the past 4-5 years, ORNL has been supporting the U.S. Department of Energy (DOE) in identifying these key technical issues, managing the collection of data to be used in issue resolution, and identifying gaps in the needed data. During this effort, ORNL subject matter experts (SMEs) have become expert in understanding what information is publicly available and what gaps in data remain. To ensure the safety of the spent fuel under normal and frequent conditions of wet and subsequent dry storage, intact fuel must be shown to: 1.Maintain fuel cladding integrity; 2.Maintain its geometry for cooling, shielding, and subcriticality; 3.Maintain retrievability, and damaged fuel with pinhole or hairline cracks must be shown not to degrade further. Where PWR (pressurized water reactor) information is utilized or referenced, justification has been provided as to why the data can be utilized for BWR fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Developing Spent Fuel Assembly for Advanced NDA Instrument Calibration - NGSI Spent Fuel Project

This report summarizes the work by Oak Ridge National Laboratory to investigate the application of modeling and simulation to support the performance assessment and calibration of the advanced nondestructive assay (NDA) instruments developed under the Next Generation Safeguards Initiative Spent Fuel (NGSI-SF) Project. Advanced NDA instrument calibration will likely require reference spent fuel assemblies with well-characterized nuclide compositions that can serve as working standards. Because no reference spent fuel standard currently exists, and the practical ability to obtain direct measurement of nuclide compositions using destructive assay (DA) measurements of an entire fuel assembly is prohibitive in the near term due to the complexity and cost of spent fuel experiments, modeling and simulation will be required to construct such reference fuel assemblies. These calculations will be used to support instrument field tests at the Swedish Interim Storage Facility (Clab) for Spent Nuclear Fuel.

Hu, Jianwei↗
DOE OSTI2013

Safety Aspects of Dry Spent Fuel Storage and Spent Fuel Management - 13559

Dry storage systems are characterized by passive and inherent safety systems ensuring safety even in case of severe incidents or accidents. After the events of Fukushima, the advantages of such passively and inherently safe dry storage systems have become more and more obvious. As with the storage of all radioactive materials, the storage of spent nuclear fuel (SF) and high-level radioactive waste (HLW) must conform to safety requirements. Following safety aspects must be achieved throughout the storage period: - safe enclosure of radioactive materials, - safe removal of decay heat, - securing nuclear criticality safety, - avoidance of unnecessary radiation exposure. The implementation of these safety requirements can be achieved by dry storage of SF and HLW in casks as well as in other systems such as dry vault storage systems or spent fuel pools, where the latter is neither a dry nor a passive system. Furthermore, transport capability must be guaranteed during and after storage as well as limitation and control of radiation exposure. The safe enclosure of radioactive materials in dry storage casks can be achieved by a double-lid sealing system with surveillance of the sealing system. The safe removal of decay heat must be ensured by the design of the storage containers and the storage facility. The safe confinement of radioactive inventory has to be ensured by mechanical integrity of fuel assembly structures. This is guaranteed, e.g. by maintaining the mechanical integrity of the fuel rods or by additional safety measures for defective fuel rods. In order to ensure nuclear critically safety, possible effects of accidents have also to be taken into consideration. In case of dry storage it might be necessary to exclude the re-positioning of fissile material inside the container and/or neutron moderator exclusion might be taken into account. Unnecessary radiation exposure can be avoided by the cask or canister vault system itself. In Germany dry storage of SF in casks fulfills both transport and storage requirements. Mostly, storage facilities are designed as concrete buildings above the ground, but due to regional constraints, one storage facility has also been built as a rock tunnel. The decay heat is always removed by natural air flow; further technical equipment is not needed. The removal of decay heat and shielding had been modeled and calculated by state-of-the-art computer codes before such a facility has been built. TueV and BAM present their long experience in the licensing process for sites and casks and inform about spent nuclear fuel management and issues concerning dry storage of spent nuclear fuel. Different storage systems and facilities in Germany, Europe and world-wide are compared with respect to the safety aspects mentioned above. Initial points are the safety issues of wet storage of SF, and it is shown how dry storage systems can ensure the compliance with the mentioned safety criteria over a long storage period. The German storage concept for dry storage of SF and HLW is presented and discussed. Exemplarily, the process of licensing, erection and operation of selected German dry storage facilities is presented. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Safety aspects of dry spent fuel storage and spent fuel management

As with the storage of all radioactive materials, the storage of spent nuclear fuel (SF) and high-level radioactive waste (HLW) must conform to safety requirements. Safety aspects like safe enclosure of radioactive materials, safe removal of decay heat, nuclear criticality safety and avoidance of unnecessary radiation exposure must be achieved throughout the storage period. The implementation of these safety requirements can be achieved by dry storage of SF and HLW in casks as well as in other systems such as dry vault storage systems or spent fuel pools, where the latter is neither a dry nor a passive system. After the events of Fukushima, the advantages of passively and inherently safe dry storage systems have become more obvious. TUV and BAM, who work as independent experts for the competent authorities, present the licensing process for sites and casks and inform about spent nuclear fuel management and issues concerning dry storage of spent nuclear fuel, based on their long experience in these fields. All safety relevant issues like safe enclosure, shielding, removal of the decay heat or behavior of cask and building under accident conditions are checked and validated with state-of-the-art methods and computer codes before the license approval. It is shown how dry storage systems can ensure the compliance with the mentioned safety criteria over a long storage period. Exemplarily, the process of licensing, erection and operation of selected German dry storage facilities is presented. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Optimizing Transmission of Acoustic Signals to Monitor Internal Conditions of Canisters for Dry Storage of Commercial Spent Nuclear Fuel

Safe storage of spent nuclear fuel (SNF) is critical to the nuclear fuel cycle and the future of nuclear energy. In the United States, SNF is stored primarily via two methods regulated by the U.S. Nuclear Regulatory Commission: wet storage in SNF pools and dry storage in dry cask storage systems (DCSSs). After about five years of cooling in spent fuel pools, the fuel assemblies are transferred into DCSSs, and the systems are filled with helium and sealed by welding. Deterioration of conditions inside of a DCSS is reflected in changes in the internal gas properties; this motivates the development of acoustic techniques to monitor internal gas properties, over extended storage periods, using sensors mounted on the exterior of the storage packages. However, a major challenge in collecting acoustic signals is the impedance mismatch between the steel canister shell and the gas. Only a small fraction of the ultrasonic signal can be transmitted through the gas medium. This paper documents experimental studies conducted on a full-scale canister mock-up to capture the gas-borne signals. Damping materials were pasted on the outside, and blocking and unblocking tests were conducted to identify the gas-borne signal. The results show that the excitation frequency plays an important role in maximizing the gas-borne signals. The gas-borne signal was successfully detected at around the theoretical time-of-flight. A high signal-to-noise ratio was achieved in the measurements. Next, the acoustic impedance matching layers were introduced, and the gas signal was drastically improved compared with that using no AIM layers.

Spent nuclear fuel (SNF), Canisters, Internal cond↗
DOE OSTI2011

Spectroscopic Monitoring of Spent Nuclear Fuel Reprocessing Streams: An Evaluation of Spent Fuel Solutions via Raman, Visible, and Near-Infrared Spectroscopy

The potential of using optical spectroscopic techniques, such as Raman and Visible/Near Infrared (Vis/NIR), for on-line process control and special nuclear materials accountability applications at a spent nuclear fuel reprocessing facility was evaluated. Availability of on-line real-time techniques that directly measure process concentrations of nuclear materials will enhance performance and proliferation resistance of the solvent extraction processes. Further, on-line monitoring of radiochemical streams will also improve reprocessing plant operation and safety. This report reviews current state of development of the spectroscopic on-line monitoring techniques for such solutions. To further examine applicability of optical spectroscopy for monitoring reprocessing solutions, segments of a spent nuclear fuel, with approximate burn-up values of 70 MWd/kgM, were dissolved in concentrated nitric acid and adjusted to varying final concentrations of HNO3. The resulting spent fuel solutions were batch-contacted with tributyl phosphate/dodecane organic solvent. The feed and equilibrium aqueous and loaded organic solutions were subjected to optical measurements. The obtained spectra showed the presence of the quantifiable Raman bands due to NO3- and UO22+ and Vis/NIR bands due to multiple species of Pu(IV), Pu(VI), Np(V), the Np(V)-U(VI) cation-cation complex, and Nd(III) in fuel solutions, justifying spectroscopic techniques as a promising methodology for monitoring spent fuel processing solutions in real-time. Quantitative evaluation of the fuel solution was performed based on spectroscopic measurements and compared to ICP-MS analysis.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Determination of the spent fuel decay heat with the VERA core simulator

VERA core simulator capabilities to predict spent fuel decay heat have been explored. The comparison with the Serpent2 and NEWT/TRITON sequence from the SCALE package, performed on the NPP Krško fuel test cases, has enabled independent verification of the successful implementation of the VERA depletion/decay calculation sequence applied on the extended nuclide set. Since the VERA is a 3-D core simulator, the focus in the rest of the paper was on the investigation of the 3-D effects and non-linearity of the soluble boron concentration averaging process, which are difficult to assess with most of available computational tools. A difference in the decay heat prediction between the 3-D and 2-D model approaches was analyzed. In addition, the impact of the Inconel grids on the decay heat was determined. Averaging the concentration of soluble boron over the fuel irradiation period has been found to be a viable approach, since it results in slight decay heat overprediction, which is considered conservative for most applications. Finally, demonstration on the Watts Bar Unit 1 cycle 1 depletion has shown formidable VERA capabilities to accurately predict spent fuel decay heat.

Kromar, Marjan↗
DOE OSTI2019

Estimation of Radiation Source Terms of Hanbit Unit3 Spent Fuel with Realistic Spent Fuel Data - 19043

Besides public acceptance, the construction of the dry storage facilities requires a sophisticated safety analysis for licensing, which needs accurate estimation of radiation source terms including the spent fuel inventories. In this work, we estimated the characteristics such as heat generations, radiotoxicities, and radioactivities of spent fuels discharged from HanBit Unit 3 using the recent realistic data of the spent fuels including detailed operational histories. The estimation was performed by writing an in-house program which automatically generates ORIGEN-S input files and processes their output files. Also, we performed the evaluation of source terms using these spent fuel assemblies. In particular, these results are compared with those obtained using a simple way with a single specific power and no cooling time between successive cycles. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

BWR Spent Nuclear Fuel Acquisition and Testing to Support DOE-NE High Burnup Spent Fuel Data Project

The Office of Spent Fuel and Waste Disposition (SFWD) within the US Department of Energy (DOE) Office of Nuclear Energy (NE) established the Spent Fuel and Waste Science and Technology (SFWST) campaign to conduct research and development (R&D) activities related to the storage, transportation, and disposal of spent nuclear fuel (SNF) and high-level radioactive waste. The SFWST program was created within SFWD to address issues of extended or long-term SNF storage and transportation. Some near-term objectives of SFWST are to use a science-based, engineering-driven approach to: Support the enhancement of the technical bases to support the continued safe and secure dry storage of SNF for extended periods; Support the enhancement of the technical bases for retrieving SNF after extended dry storage; Support the enhancement of the technical bases for transporting high burnup (HBU) fuel and transporting low burnup fuel and HBU fuel after dry storage DOE-NE, in partnership with the Electric Power Research Institute, developed the High Burnup Spent Fuel Data Project to perform a large-scale demonstration and laboratory-scale testing of HBU pressurized water reactor (PWR) fuels (exceeding 45 gigawatt-days per metric ton of uranium [GWd/MTU]). Under this project, 25 sister rods—which are rods that have the same design, power histories, and other characteristics—were removed from assemblies at the North Anna Nuclear Power Station and sent to Oak Ridge National Laboratory (ORNL) in January 2016. ORNL performed detailed nondestructive examination (NDE) on all 25 rods. The NDE consisted of visual examinations, gamma and neutron scanning, profilometry and rod length measurements, and eddy current examinations. After completing the NDE, 10 of the sister rods were delivered to Pacific Northwest National Laboratory (PNNL) in a NAC International, Inc. legal-weight truck cask in September 2018 for destructive examination (DE). To date, SFWD work has focused on the PWR fuel that is part of the Sister Rod Test program. No boiling water reactor (BWR) fuel has been tested in the program, and the data needs that were identified for the PWR fuel have not been collected for BWR fuel. The goal to obtain six to nine BWR rods and test them at ORNL will support closing this important data gap. BWR fuel comprises approximately 56% of the total fuel assemblies currently in storage at nuclear power plants in the United States. BWR nuclear fuel and cladding designs and manufacturing are significantly different from PWRs. Differences include the following: BWR fuel pellets are larger than PWR pellets; Variations of Zircaloy-2 (including liners) are used instead of the Zircaloy-4 cladding materials used in PWRs; Clad manufacturing and stress-relief processes are different between PWRs and BWRs; The fuel rod dimensions are different because larger rod diameters and thicker cladding are used in BWRs; BWR fuel typically has lower internal rod pressures and sees vastly different operating conditions than PWR fuel (i.e., two-phase flow); BWR assemblies are “canned,” meaning each assembly is surrounded by a metal fuel channel; BWR cladding is often composed of an inner pure Zr liner that has widely different mechanical properties than the Zircaloy-2 alloy and exhibits a stronger affinity for hydrogen; The BWR SNF generally has more total hydrogen in the cladding/liner than typical PWR fuel; The construction of the PWR and BWR assemblies is vastly different; BWR rods are solidly attached to the assembly nozzles and experience a much different vibration and shock load than PWR rods, which are “floating” within a grid system attached to guide tubes, and the rods sit loosely on the bottom end plates. These numerous differences will affect the way the BWR SNF responds under dry storage preparation processes (e.g., vacuum drying) and during transportation. The results collected in the PWR experimental program must be compared with a subset of similar data collected on BWR SNF to establish a technical basis for whether the larger PWR database is sufficient to bound the BWR SNF end-of-life conditions as is currently assumed for several fuel/clad properties. Changes that occur in both fuel types at HBU could exacerbate any mechanical property differences. As the fuel burnup increases, several changes occur that might affect the performance of the fuel, cladding, and assembly hardware in storage and transportation. These changes include increased cladding corrosion layer thickness, increased cladding hydrogen content, increased cladding creep strains, increased fission gas release, and the formation of the HBU structure at the surface of the fuel pellets. The Nuclear Regulatory Commission (NRC) limits the current maximum rod-averaged burnup to 62 GWd/MTU due to these changes and the lack of data at higher burnups.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗