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All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

FRAPCON analysis of cladding performance during dry storage operations

There is an increasing need in the U.S. and around the world to move used nuclear fuel from wet storage in fuel pools to dry storage in casks stored at independent spent fuel storage installations (ISFSI) or interim storage sites. The NRC limits cladding temperature to 400°C while maintaining cladding hoop stress below 90 MPa in an effort to avoid radial hydride reorientation. An analysis was conducted with FRAPCON-4.0 on three modern fuel designs with three representative used nuclear fuel storage temperature profiles that peaked at 400 °C. Results were representative of the majority of U.S. LWR fuel. They conservatively showed that hoop stress remains below 90 MPa at the licensing temperature limit. Results also show that the limiting case for hoop stress may not be at the highest rod internal pressure in all cases but will be related to the axial temperature and oxidation profiles of the rods at the end of life and in storage.

Dry Storage↗
DOE OSTI2023

Investigating microstructural evolution in SolidStir™ extruded oxide dispersion strengthened 14YWT alloy fuel cladding tube

Structural components for extreme environments require advanced materials and manufacturing processes. One such advanced material is oxide dispersion strengthened (ODS) Fe–14 wt.%Cr–3W–0.4Ti–0.3Y 2 O 3 (14YWT), which is developed for components to be used in Gen IV and fusion nuclear reactors. However, the conventional manufacturing processes to produce components, such as fuel cladding tubes, either do not retain the desired microstructural attributes or are costly and less efficient. As a result, a more attractive option is a novel manufacturing process developed on the principle of friction stir welding or processing (FSW/P) and commercially referred to as SolidStir™ Extrusion (SSE). Here this study used the SSE technique to study the manufacturing and microstructural evolution of a fuel cladding tube made of ODS 14YWT alloy. The cladding extrusion by the SSE technique involved the use of ball-milled 14YWT powders with Y, Ti, and O in the solid solution and the use of a specially designed W-25Re-Hf tool for consolidation and extrusion of the powders. A scanning electron microscope (SEM), electron backscattered diffraction (EBSD), and transmission electron microscope (TEM) were utilized for microstructural characterization. A Keyence microscope captured macrostructural photographs of the extruded tube. EBSD examination of the extruded tube on both transverse and longitudinal cross-sections showed the presence of dynamically recrystallized grains and revealed that the average grain size of the transverse cross-section was smaller than that of the longitudinal cross-section. The texture was weak and consisted of some amount of shear texture. The presence of nano-oxide clusters or precipitates of Y, Ti, and O (pyrochlores) in the extruded tube was determined by TEM. Aging heat treatment caused the average precipitate size to decrease and the density to increase compared to the as-processed condition. The results indicated that SSE is a viable tool for manufacturing fuel cladding tubes with the desired microstructural attributes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Update on the Investigation of Commercial Drying Cycles Using the Advanced Drying Cycle Simulator

The purpose of this report is to document updates on the apparatus to simulate commercial vacuum drying procedures at the Nuclear Energy Work Complex at Sandia National Laboratories. Validation of the extent of water removal in a dry spent nuclear fuel storage system based on drying procedures used at nuclear power plants is needed to close existing technical gaps. Operational conditions leading to incomplete drying may have potential impacts on the fuel, cladding, and other components in the system during subsequent storage and disposal. A general lack of data suitable for model validation of commercial nuclear canister drying processes necessitates well-designed investigations of drying process efficacy and water retention. Scaled tests that incorporate relevant physics and well-controlled boundary conditions are essential to provide insight and guidance to the simulation of prototypic systems undergoing drying processes. This report documents a new test apparatus, the Advanced Drying Cycle Simulator (ADCS). This apparatus was built to simulate commercial drying procedures and quantify the amount of residual water remaining in a pressurized water reactor (PWR) fuel assembly after drying. The ADCS was constructed with a prototypic 17×17 PWR fuel skeleton and waterproof heater rods to simulate decay heat. These waterproof heaters are the next generation design to heater rods developed and tested at Sandia National Laboratories in FY20. This report describes the ADCS vessel build that was completed late in FY22, including the receipt of the prototypic length waterproof heater rods and construction of the fuel basket and the pressure vessel components. In addition, installations of thermocouples, emissivity coupons, pressure and vacuum lines, pressure transducers, and electrical connections were completed. Preliminary power functionality testing was conducted to demonstrate the capabilities of the ADCS. In FY23, a test plan for the ADCS will be developed to implement a drying procedure based on measurements from the process used for the High Burnup Demonstration Project. While applying power to the simulated fuel rods, this procedure is expected to consist of filling the ADCS vessel with water, draining the water with applied pressure and multiple helium blowdowns, evacuating additional water with a vacuum drying sequence at successively lower pressures, and backfilling the vessel with helium. Additional investigations are expected to feature failed fuel rod simulators with engineered cladding defects and guide tubes with obstructed dashpots to challenge the drying system with multiple water retention sites.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Oxide Morphology of C26M at 300 - 600 °C

Iron-chromium-aluminum (FeCrAl) alloys are of interest to the nuclear materials community due to their resistance to high temperature steam oxidation under accident conditions. The present work investigates oxide formation at temperatures relevant to light water reactor cladding operation following extended aging to assess growth kinetics, chemical composition, and microstructure of oxide formation on an Oak Ridge National Laboratory developed Fe-12Cr-6Al- 2Mo-0.2Si-0.03Y alloy (designated ‘C26M’). Aging treatments were performed for 100-2000 hours in stagnant air at 300, 400, 500, and 600 °C, respectively. Oxide growth behavior under the investigated conditions follows a logarithmic time dependence. When the oxidization temperature is 400 °C or below, the oxide is amorphous. At 500 °C, crystalline α-Al 2 O 3 oxide film develops and the correlated logarithmic rate constant decreases significantly, indicating enhanced oxidation resistance of the formed oxide film. In addition, Mo segregation at grain boundaries has been observed when the aging temperature exceeds 500 °C. The results of this study can be viewed as an upper bounding result for potential oxide coarsening on the cladding inner surface during reactor operation.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Molybdenum Sleeve Experiments in Fully-Reflected Water-Moderated Triangular-Pitched U(6.90)O 2 Fuel Rod Lattices (1.55 cm Pitch)

The US Department of Energy (DOE) Nuclear Energy Research Initiative funded the design and construction of the Seven Percent Critical Experiment (7uPCX) at Sandia National Laboratories. The start-up of the experiment facility and the execution of the experiments described here were funded by the DOE Nuclear Criticality Safety Program. The 7uPCX is designed to investigate critical systems with fuel for light water reactors in the enrichment range above 5 % 235 U. The 7uPCX assembly is a water-moderated and -reflected array of aluminum-clad U(6.90 %)O 2 fuel rods. Other critical experiments performed in the 7uPCX assembly are documented in LEU-COMP-THERM-078, LEU-COMP-THERM-080, LEU-COMP-THERM-096, LEUCOMP-THERM-097, LEU-COMP-THERM-101, and LEU-COMP-THERM-102. The purpose of these experiments was to measure the effects of molybdenum in nearly-critical systems. The molybdenum was introduced into the fuel arrays as tubular sleeves that surrounded some of the fuel rods in the fuel arrays measured. Four hundred molybdenum tubes nominally 12.7 mm outside diameter, 498 mm long, with 0.762 mm wall thickness were provided for the experiments by the Institut de Radioprotection et de Sûreté Nucléaire (IRSN). Small polyethylene adapters at each end of the tubes were used to center each tube on a fuel rod in the assembly. The critical experiments were done using a set of triangular-pitched grid plates fabricated for these experiments. The grid plate set accommodated a fuel array of a total of 1261 fuel rod positions on a pitch of 0.610 in (1.5494 cm) in a series of 20 hexagonal rings surrounding the central fuel rod. The fuel used in these experiments was fabricated using unirradiated 6.90 % enriched UO 2 fuel pellets from fuel elements designed to be used in the internal nuclear superheater section of the Pathfinder boiling water reactor operated in South Dakota by the Northern States Power Company in the 1960s. The fuel elements were obtained from The Pennsylvania State University where they had been stored for many years. The fuel pellets in those fuel elements were removed from the original Incoloy cladding and reclad in 3003 aluminum tubes and end caps for use in the experiments reported here. The five critical experiments in this series were performed in August through December 2022, in the Sandia Critical Experiments (SCX) at the Sandia Pulsed Reactor Facility. Case 1 had no molybdenum sleeves, Case 2 had 208 molybdenum sleeves clustered at the center of the array, Case 3 had 397 molybdenum sleeves clustered at the center of the array, Case 4 had 175 molybdenum sleeves in the central position and in five alternating hexagonal rings, and Case 5 had 331 molybdenum sleeves in the central position and in seven alternating hexagonal rings. All five critical experiments are judged to be acceptable as benchmark experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Waveguides for neurostimulation in the cochlea

Introduction Optical stimulation has been suggested for neural stimulation to improve cochlear implants. Light allows for more spatially selective activation of neuron populations than electrical current, offering more independent frequency bands along the spiral ganglion. These bands are available to encode acoustic information with anticipated better frequency resolution, improving cochlear implant user performance in noisy listening environments, tonal languages, and music perception. Methods Optical cochlear implants (oCIs) can deliver light either directly via small emitters within the cochlea or via waveguides from external optical sources. We investigated three waveguide designs made from OrmoComp ® , a polymer that cures through ultraviolet (UV) radiation. Waveguides were fabricated via injection molding and coated using dip-coating or thermal reflow, or through aspiration of OrmoComp ® into polyimide tubing that served as the cladding of the waveguide. The choice of fabrication technique directly determined the waveguides' total diameter: thermal reflow yielded ≈940 μm, dip-coating produced ≈306 μm, and aspiration resulted in ≈132 μm core diameter waveguides. Given the human cochlea's small size, we focused on analyzing the 306-μm and 132-μm waveguides, evaluating their optical performance (propagation and bending losses) and mechanical properties (bending stiffness and insertion forces). Furthermore, we evaluated some of these designs in in-vivo guinea pigs experiments. Results For the 100-μm core diameter waveguides, the propagation losses were 12.34 ± 1.26, 1.18 ± 0.88, 1.49 ± 0.58, and 3.43 ± 0.68 dB/cm at 534, 1,375, 1,460, and 1,550 nm, respectively. The respective bending losses at a 2 mm radius of curvature were 5.50 ± 1.32, 0.56 ± 0.26, 0.79 ± 0.18, and 0.64 ± 0.23 dB, and at 1 mm 8.54 ± 1.30, 2.05 ± 0.84, 2.11 ± 0.50, and 1.44 ± 0.37 dB. The bending stiffness of a 1 mm segment of the 100-μm-diameter waveguides was 18.9 ± 2.2 N/m. Insertion forces for the 100-μm-diameter waveguides into an acrylic human-size scala tympani model were < 25 mN. For the waveguides with 306 and 940 μm total diameter, the propagation losses ranged between 0.43 and 2.40 dB/cm at 534, 680, 1,375, and 1,550 nm, between 2.19 and 3.78 dB/cm at 450 and 1,460 nm. Bending losses for 360 degrees at 1,375 nm were 5.0, 2.4, and 0.46 for a bending radius of 2.5-, 3-, and 4-mm. Discussion Our study demonstrated that the polymer OrmoComp ® is suitable for fabricating waveguides to transmit near-infrared radiation. In-vivo experiments showed optically evoked auditory responses originating from optical stimulation in the guinea pigs' first cochlear turn while radiation was delivered. Incoming experiments will focus on evaluating long-term performance of these waveguides in guinea pigs and cats. This study will be designed to provide insights into the waveguides' performance and biocompatibility over extended periods, essential for their potential clinical application in future oCIs.

Kim, Matthew↗
DOE OSTI2017

Accident tolerant FeCrAl fuel cladding: current status towards commercialization

FeCrAl alloys are rapidly becoming a mature concept for accident tolerant fuel applications. The FeCrAl material class has shown excellent oxidation resistance in high-temperature steam environments, a key aspect of any accident tolerant cladding concept, while also being corrosion resistant, stress corrosion cracking (SCC) resistant, irradiation-induced swelling resistant, weldable, and formable. Current research efforts are focused on design, development and commercial scaling of advanced FeCrAl alloys including large-scale thin-walled seamless tube production followed by a broad spectrum of degradation evaluations in both normal and off-normal conditions. Included in this discussion will be theoretical analysis of the alloying principles and rules, alloy composition design, and overview of the most recent empirical database on possible degradation phenomena for FeCrAl alloys. The results are derived from extensive in-pile and out-of-pile experiments and form the basis for near-term deployment of a lead-test rod and/or assembly within a commercially operating nuclear power plant.

Field, Kevin G.↗
DOE OSTI2018

Water-Moderated U(4.31)O 2 Fuel Rod Lattices Containing Rhodium Foils

The US Department of Energy Nuclear Energy Research Initiative (NERI) funded the Burnup Credit Critical Experiment (BUCCX) at Sandia National Laboratories. The BUCCX was designed to investigate the effect of fission product materials on critical systems. The BUCCX assembly was a water-moderated and -reflected array of Zircaloy-clad triangular-pitched U(4.31%)0 2 fuel elements. Some of the fuel elements could be opened to allow placement of experiment materials between the fuel pellets in the element. The ten BUCCX critical experiments reported here test the effect of the fission product rhodium on the assembly. The calculated reactivity worth of the rhodium in the experiments ranged from 0% for cases with no rhodium to a maximum of 3.5% of k eff .

103-1-↗
DOE OSTI2019

Titanium and/or Aluminum Sleeve Experiments in Fully-Reflected Water-Moderated U(4.31)O 2 Fuel Rod Lattices with 2.8 cm Pitch.

The US Department of Energy Nuclear Energy Research Initiative (NERI) funded the Bumup Credit Critical Experiment (BUCCX) at Sandia National Laboratories. The BUCCX was designed to investigate the effect of fission product materials on critical systems. The BUCCX assembly is a water-moderated and -reflected array of Zircaloy-clad triangular-pitched U(4.31)02 fuel elements. The original BUCCX experiments with rhodium are evaluated in LEU-COMP-THERM-079. In the experiments here, sets of up to 60 experiment titanium and aluminum sleeves with nominal outside diameter of 1 in (2.54 cm), wall thickness of 0.035 in (0.0889 cm), and length of 19.60 (49.784 cm) were fabricated. The sleeves are approximately the same length as the fueled section of the fuel elements and have an inner diameter that is 0.421 in (1.0693 cm) larger than the fuel elements. This allows for each sleeve to be centered around a fuel element between the grid plates within the array. Configurations differ by the number and location of sleeves. The seventeen BUCCX critical experiments reported here compare the effects of the titanium and aluminum sleeves on nearly critical fuel assembly arrays.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Update on the Simulation of Commercial Drying of Spent Nuclear Fuel

The purpose of this report is to document improvements in the simulation of commercial vacuum drying procedures at the Nuclear Energy Work Complex at Sandia National Laboratories. Validation of the extent of water removal in a dry spent nuclear fuel storage system based on drying procedures used at nuclear power plants is needed to close existing technical gaps. Operational conditions leading to incomplete drying may have potential impacts on the fuel, cladding, and other components in the system. A general lack of data suitable for model validation of commercial nuclear canister drying processes necessitates additional, well-designed investigations of drying process efficacy and water retention. Scaled tests that incorporate relevant physics and well-controlled boundary conditions are essential to provide insight and guidance to the simulation of prototypic systems undergoing drying processes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Response of a Pressurized Water Reactor Dashpot Region to Commercial Drying Cycles

The purpose of this report is to document updates to the simulation of commercial vacuum drying procedures at the Nuclear Energy Work Complex at Sandia National Laboratories. Validation of the extent of water removal in a dry spent nuclear fuel storage system based on drying procedures used at nuclear power plants is needed to close existing technical gaps. Operational conditions leading to incomplete drying may have potential impacts on the fuel, cladding, and other components in the system. A general lack of data suitable for model validation of commercial nuclear canister drying processes necessitates additional, well-designed investigations of drying process efficacy and water retention. Scaled tests that incorporate relevant physics and well-controlled boundary conditions are essential to provide insight and guidance to the simulation of prototypic systems undergoing drying processes. This report documents testing updates for the Dashpot Drying Apparatus (DDA), an apparatus constructed at a reduced scale with multiple Pressurized Water Reactor (PWR) fuel rod surrogates and a single guide tube dashpot. This apparatus is fashioned from a truncated 5×5 section of a prototypic 17×17 PWR fuel skeleton and includes the lowest segment of a single guide tube, often referred to as the dashpot region. The guide tube in this assembly is open and allows for insertion of a poison rod (neutron absorber) surrogate.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Gap Analysis of Material Properties Data for Ferritic/Martensitic HT-9 Steel

The US Department of Energy (DOE), Office of Nuclear Energy (NE), is supporting the development of an ASME Code Case for adoption of 12Cr-1Mo-VW ferritic/martensitic (F/M) steel, commonly known as HT-9, primarily for use in elevated temperature design of liquid-metal fast reactors (LMFR) and components. In 2011, Los Alamos National Laboratory (LANL) nuclear engineering staff began assisting in the development of a small modular reactor (SMR) design concept, previously known as the Hyperion Module, now called the Gen4 Module. LANL staff immediately proposed HT-9 for the reactor vessel and components, as well as fuel clad and ducting, due to its superior thermal qualities. Although the ASME material Code Case, for adoption of HT-9 as an approved elevated temperature material for LMFR service, is the ultimate goal of this project, there are several key deliverables that must first be successfully accomplished. The most important key deliverable is the research, accumulation, and documentation of specific material parameters; physical, mechanical, and environmental, which becomes the basis for an ASME Code Case. Time-independent tensile and ductility data and time-dependent creep and creep-rupture behavior are some of the material properties required for a successful ASME Code case. Although this report provides a cursory review of the available data, a much more comprehensive study of open-source data would be necessary. This report serves three purposes: (a) provides a list of already existing material data information that could ultimately be made available to the ASME Code, (b) determines the HT-9 material properties data missing from available sources that would be required and (c) estimates the necessary material testing required to close the gap. Ultimately, the gap analysis demonstrates that certain material properties testing will be required to fulfill the necessary information package for an ASME Code Case.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Loss-of-Coolant Accident Analysis of a High-Burnup Pressurized Water Reactor Core Design Using Gadolinia-Doped UO 2

High-burnup and extended enrichment fuels are of interest for extending the cycle lengths of pressurized water reactors from 18 to 24 months. Changes to the fuel design and core loading scheme have potentially significant safety implications due to power distribution effects, reduced thermal conductivity, and/or increased plenum pressures due to additional burnable poison loadings. Additionally, higher burnups result in increased material degradation and risk of fuel fragmentation, relocation, and dispersal (FFRD). A representative 24-month core design using gadolinia-doped UO 2 was analyzed for performance under large-break (LB) loss-of-coolant accident (LOCA) conditions using PARCS, RELAP5-3D, and BISON. Furthermore, all considered acceptance criteria were met, with no cases exceeding the 1477 K maximum cladding temperature or the post-quench ductility oxidation limit of 17% equivalent cladding reacted. Full-core FFRD susceptibility was estimated to be approximately 455 kg, though high uncertainties exist with current approaches for computing susceptibility. Undoped fuel rods are more likely to be limiting due to higher linear heat rates. Relatively high burnup and linear heat rate rods located in second batch assemblies are of greatest safety significance during LB LOCA for this high-burnup core design.

High burnup↗
DOE OSTI2016

Finite element analysis of the PWR control rod absorber thermal evaluation

The RCCAs (Rod Cluster Control Assembly) are part of the reactor control systems and are designed to adjust the reactivity by neutron absorption. In PWR, a RCCA is composed of a spider holding 24 control rods. During operation, the absorber stacks are heated up by neutron and γ-ray radiation. This paper describes the PWR control rod absorber temperature analysis model in details, derives the theoretical equations, and presents a FEA model for control rod absorber thermal evaluation. The FEA model is based on finite element analysis and on 3 major assumptions: 1) steady-state condition, 2) the thermal conductivities of absorber, cladding and gas are assumed to be constant, and 3) thermal expansion and radiation are neglected. The control rod of an AFA 3G fuel assembly in Ningde NPP unit 1 and 2 in steady state on condition I operation was used as a case to qualify the FEA method. A special software named CRABE, designed and qualified by AREVA, was used to do the same analysis for comparison. Several conclusions can be drawn. First, compared with traditional methods, FEA is more convenient with higher accuracy and can greatly simplify the calculation process. In addition, the FEA software is always ease to use, with a friendly interface and a more intuitive display of results. Secondly, the FEA model was qualified though a practical case calculated in different ways. Thirdly, when it comes to a special structure such as an eccentric absorber or an oval cladding, the FEA model has a great applicability, while traditional methods are difficult to use.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Efficient computational search for lanthanide-binding additive dopants for advanced U-Zr based fuels

A rational design strategy combining theory and experiment is highly desirable for the development of U-Zr based fuels that are resistant to lanthanide-induced fuel-cladding chemical interaction (FCCI). Here, we performed rapid computational screening of FCCI-mitigating (via strong lanthanide-binding) elements across the periodic table by utilizing three simple criteria and density functional theory (DFT) calculated total energies of relevant elements and compounds retrieved from high-throughput DFT databases. In addition to successfully identifying previously demonstrated additives, our search leads to the discovery of Bi as a highly effective element among all 68 elements screened, and its effectiveness was experimentally confirmed in this study.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Hot Isostatic Press Can Optimization for Aluminum Cladding of U-10Mo Reactor Fuel Plates: FY12 Final Report and FY13 Update

Currently, the proposed processing path for low enriched uranium – 10 wt. pct. molybdenum alloy (LEU-10Mo) monolithic fuel plates for high power research and test reactors includes hot isostatic pressing (HIP) to bond the aluminum cladding that encapsulates the fuel foil. Initial HIP experiments were performed at Idaho National Laboratory (INL) on approximately ¼ scale “mini” fuel plate samples using a HIP can design intended for these smaller experimental trials. These experiments showed that, with the addition of a co-rolled zirconium diffusion barrier on the LEU-10Mo alloy fuel foil, the HIP bonding process is a viable method for producing monolithic fuel plates. Further experimental trials at Los Alamos National Laboratory (LANL) effectively scaled-up the “mini” can design to produce full-size fuel prototypic plates. This report summarizes current efforts at LANL to produce a HIP can design that is further optimized for higher volume production runs. The production-optimized HIP can design goals were determined by LANL and Babcock & Wilcox (B&W) to include maintaining or improving the quality of the fuel plates produced with the baseline scaled-up mini can design, while minimizing material usage, improving dimensional stability, easing assembly and disassembly, eliminating machining, and significantly reducing welding. The initial small-scale experiments described in this report show that a formed-can approach can achieve the goals described above. Future work includes scaling the formed-can approach to full-size fuel plates, and current progress toward this goal is also summarized here.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Extended Storage of High Burn-up Fuel Cask R and D Project - 17145

Under contract by EPRI and DOE, AREVA Federal Services LLC (AFS) is leading a team consisting of AREVA Inc. and Dominion Virginia Power (Dominion) to develop and conduct a full-scale confirmatory Research project with HBU fuel. A test plan was developed using an existing TN Americas (TN) TN-32 bolted storage cask with a modified, specially instrumented lid to be engineered and then loaded with multiple cladding types of HBU pressurized water reactor (PWR) UNF from Dominion's North Anna Power Station (this modified cask is identified as TN-32B). Seven specially designed thermocouple lances, each containing nine detectors, will be inserted into the designated fuel assembly guide tubes through the modified cask lid. The loaded cask with the monitoring system will be placed into dry storage on the existing North Anna Independent Spent Fuel Storage Installation (ISFSI) pad for a period of extended monitoring. In addition, design modifications to the lid have been incorporated to permit periodic gas sampling of the cask internals during the storage period on the ISFSI pad. The fuel selected for this research project ranges in average assembly burnup between 50 GWD/MTU to 67 GWD/MTU. The instrumentation used to collect the thermal data is very similar to in-core thermocouple detectors used at nuclear power plants and is considered very reliable for this application. Prior to cask loading at North Anna, a series of 'fit up' tests will be performed at the fabricator to ensure proper alignment, fit and function of the modified cask and related instrumentation and equipment. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Extended Storage of High Burn-up Fuel Cask R and D Project - 16591

Under contract by EPRI and DOE, AREVA Federal Services LLC (AFS) is leading a team consisting of AREVA Inc. and Dominion Virginia Power (Dominion) to develop and conduct a full-scale confirmatory Research project with HBU fuel. A test plan was developed using an existing AREVA Transnuclear (TN) TN-32 bolted storage cask with a modified, specially instrumented lid to be engineered and then loaded with multiple cladding types of HBU pressurized water reactor (PWR) UNF from Dominion's North Anna Power Station (this modified cask is identified as TN-32B). Seven specially designed thermocouple lances, each containing nine detectors, will be inserted into the designated fuel assembly guide tubes through the modified cask lid. The loaded cask with the monitoring system will be placed into dry storage on the existing North Anna Independent Spent Fuel Storage Installation (ISFSI) pad for a period of extended monitoring. In addition, design modifications to the lid have been incorporated to permit periodic gas sampling of the cask internals during the storage period on the ISFSI pad. The fuel selected for this research project ranges in average assembly burnup between 50 GWD/MTU to 67 GWD/MTU. The instrumentation used to collect the thermal data is very similar to in-core thermocouple detectors used at nuclear power plants and is considered very reliable for this application. Prior to cask loading at North Anna, a series of 'fit up' tests will be performed at the fabricator to ensure proper alignment, fit and function of the modified cask and related instrumentation and equipment. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗