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All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Engineering Scale Drying of Aluminum-Clad Spent Nuclear Fuel: Experiment Report

Engineering-scale tests were performed to evaluate the relative effectiveness of forced gas and vacuum-drying processes on aluminum-clad spent nuclear fuel (ASNF) in preparation for extended dry storage. Drying models were developed to improve confidence in the likely range of conditions at the start of dry storage and to understand the factors most significant to each process. Testing was performed at the Holtec International Forced Helium Dehydration training facility in Camden, New Jersey, by students from the University of South Carolina in collaboration with Idaho National Laboratory. A full-size Type 1a basket, with a full-diameter vessel configuration (based on the Department of Energy standardized canister design), was used in combination with a 10 assembly surrogate ASNF arrangement to provide a proxy for geometry. A resistance heater incorporated in one assembly simulated the influence of 100 W decay heat. This instrumented chamber allowed testing of the removal of bulk water, physisorbed water, and chemisorbed water over a range of operating parameters with either drying process. Thick oxide films were grown on aluminum test plates to mimic the ASNF surface chemistry. These one-use test plates were housed within select assemblies during drying. Post-test analysis of these plates gave a measure of the moisture removed during each test, which was compared to a control. Instrumentation provided detailed thermal, humidity, and gas or pressure data relevant to each test. Model simulations generally compared favorably to instrument data for the tests. While both processes were able to remove bulk water, removal of chemisorbed water was shown to depend heavily on local temperature. To the disadvantage of vacuum drying, operations approaching or exceeding 220°C achieved consistently more dehydroxylation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Drying Wet Stored and Corroded Magnox Fuel for Interim Dry Storage - 20187

To investigate the feasibility of drying corroded Magnesium alloy clad nuclear fuel, the corrosion products of inactive and unirradiated Magnox simulant dry corrosion product and dried corroded Magnox sludge were characterised by X-ray diffraction and thermogravimetric analysis. XRD identified brucite in the dried corrosion product and both brucite and hydromagnesite in the sludge. TGA identified mass loss on dehydration of ∼9.1% up to 250 deg. C and 27.9% from dehydroxylation up to 440 deg. C, with total mass loss of 41.9% up to 800 deg. C for the dried corrosion product. The sludge TGA showed 8.3% mass loss up to 250 deg. C from dehydration, 7.5% up to 330 deg. C from dehydroxylation and 31.4% mass loss up to 47.2% for decarbonation, with total mass loss of 49.2% up to 800 deg. C. Water removal up to 1.8 g was performed by cold vacuum drying (40 deg. C-120 deg. C) on a ≅14.8 g (wet) corroded Magnox sample. The process was monitored by observing pressure, dew point, temperature and gas flow changes supported by measuring the sample mass loss as water is removed. From these tests, it was observed that the dried corroded Magnox displayed some hygroscopic capacity which has implications for water retention following vacuum drying and the length of exposure required to achieve the desired dryness. Whilst at higher temperatures the water was removed faster, the increased temperature also risks the fuel cladding undergoing unwanted chemical reactions with the residual water during the process. However, at lower temperatures the achievable levels of dryness were reduced. Therefore, from these observations this work has identified that there is a temperature balance that may be necessary to optimise the drying process with respect to allowing the greatest level of dryness whilst restricting unwanted chemical reactions. (authors)

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

Probing Pore Size and Connectivity in Porous Silicas Using 13C MAS NMR Spectroscopy of Supercritical Methane

Measuring the pore size and pore-size distributions and exploring the fluid-exchange dynamics between different types of pores in porous materials remains a significant experimental challenge but is critical to understanding catalysis, chromatography, nutrient cycling, and a whole range of geochemical phenomena, including shale gas and tight gas extraction. Here, we present the results of 1D 13 C NMR and 2D exchange spectroscopy (EXSY) NMR investigations of a porous silica using supercritical methane (scCH 4 ) as a direct probe of pore size and fluid exchange between pore types. The results show that the 13 C chemical shift of scCH 4 adsorbed in nanometer-scale silica pores becomes more negative with increasing pore diameter, in agreement with trends reported for gas hydrates, zeolites, MOFs, and clays and other microporous (<10 nm pores) geochemical materials. These 13 C chemical shifts follow a natural log–linear trend with pore size in the vacuum-dried porous silicas studied here, allowing one to predict pore size based on the 13 C chemical shift in dry silica nanopores. The EXSY and 1D 13 C NMR results both show that CH 4 exchanges between pore and bulk fluid environments over rate scales from 0.01 to 1 kHz in vacuum-dried samples and that exchange dynamics must be considered when interpreting 13 C NMR of pore-adsorbed CH 4 . Introducing H 2 O to the system causes the 13 C chemical shift to become more negative with increasing H 2 O content as a result of H 2 O preferentially filling small pores, forcing CH 4 to occupy larger pores that are better connected to the bulk environment. Likewise, the 1D 13 C data show a decreased distribution of exchange rates between pore-adsorbed and bulk CH 4 in the presence of H 2 O.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Fission Product Gas Monitoring During Fuel Drying Operations - 20094

The goal of the project was to devise and implement a method of accurately qualifying any noble gas emitted during fuel drying operations (part of the dry fuel storage process) in order to provide the data needed to validate the off-site dose calculations. The imputes of the project was the NRC publishing Information Notice 18-01, 'Noble Fission Gas Releases During Spent Fuel Cask Loading Operations' in February 2018. The information notice describes some industry events that occurred during vacuum drying operations as part of a dry fuel campaign. The solution was devised by Fort Calhoun Station staff in partnership with Mirion Technologies technical staff. The monitoring was accomplished by directing all the effluent from the fuel storage cask vacuum drying machine through a sample chamber containing a compact ISOCS characterized CZT based gamma spectrometer connected to a Mirion Data Analyst. This resulted near real-time quantification of the radionuclides of concern. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2022

The Mechanical Response of High Burnup 17x17 PWR Fuel Rods Under Bending

As a part of the US Department of Energy Office of Nuclear Energy High Burnup Spent Fuel Data Project, Oak Ridge National Laboratory has performed 4pt-bend tests on a series of 17x17 PWR fuel rods that were discharged with rod average burnups between 50 and 60 GWd/MTU. The rods include four different kinds of fuel cladding: standard Zircaloy-4 (Zirc-4), low-tin (LT) Zirc-4, ZIRLO, and M5. Six rods of each type (combining Zirc-4 and LT Zirc-4 as one type) were tested to failure at room temperature and 200 oC. Three of each type experienced a full-length heat treatment cycle from RT to 400 oC to RT with an 8 hr soak at 400 oC to simulate the expected limiting vacuum drying condition, which potentially induces hydride reorientation and causes some annealing of irradiation damage in the cladding. Overall, the fuel rods show significant capacity to deflect in bending and exhibit non-linear deformation before failing. The calculated failure strains range from 1.5% to 3.8%. The simulated vacuum drying heat treatment had the largest effect on fuel rods with ZIRLO and M5 cladding. In both cases the heat treatment caused increases in the plastic strain at failure, suggesting some annealing of irradiation damage is occurring. The finite element modeling of the 4pt bending test assume the pellet is a collection of pellet wedge pieces. The results for the Zircaloy-4 rods suggest that the pellets are bonded to the cladding and friction controls the relative motion of the pellet wedge pieces.

Cantonwine, Paul↗
DOE OSTI2024

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

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

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

Extending the family of reduced [Mn 12 O 12 (O 2 CR) 16 (H 2 O) x ] n- complexes, and their sensitivity to environmental factors

Here, the family of reduced [Mn 12 O 12 (O 2 CR) 16 (H 2 O) 4 ] n- (n = 1, 2) single-molecule magnets (SMMs) has been expanded with the synthesis of new members using I - as a stoichiometric one-electron reducing agent of the neutral n = 0 complexes. Electrochemical data, 1 H NMR spectra of the [Mn 12 O 12 (O 2 CCH 2 Cl) 16 (H 2 O) 3 ] 0,-,2- salts, and the crystal structures of two complexes, (PPh 4 )[Mn 12 O 12 (O 2 CCHCl 2 ) 16 (H 2 O) 4 ] and (PPh 4 ) 2 [Mn 12 O 12 (O 2 CCH 2 Cl) 16 (H 2 O) 3 ] have been obtained. For (PPh 4 ) 2 [Mn 12 O 12 (O 2 CCH 2 Cl) 16 (H 2 O) 3 ], the data confirm an S = 10 ground state, and magnetization vs dc field scans on a pristine crystal exhibit hysteresis loops possessing steps due to quantum tunneling of magnetization (QTM). In addition, two forms of the cluster have been identified in the crystal with distinctly different relaxation barriers, with the faster-relaxing form being the majority one in pristine crystals from mother liquor, and the slower-relaxing one becoming the majority one in vacuum-dried samples. Similar behavior is seen for (PPh 4 )[Mn 12 O 12 (O 2 CCHCl 2 ) 16 (H 2 O) 4 ] with an S = 19 / 2 ground state, which shows three forms to be present in the hysteresis loops of a pristine crystal, all converting to a slow-relaxing form on vacuum-drying. High-frequency EPR spectra of the latter confirm a high axial anisotropy with $\mathcal{D}$ = -0.477 cm -1 and rationalize its effective relaxation barrier U eff = 57 K. The combined work emphasizes a high sensitivity of such ultra-small nanomagnets to environmental influences, affecting their properties significantly, especially their quantum properties that are so important to many potential new 21st century applications.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2022

Solvent-Free Manufacturing of Lithium-Ion Battery Electrodes via Cold Plasma

Slurry casting has been used to fabricate lithium-ion battery electrodes for decades, which involves toxic and expensive organic solvents followed by high-cost vacuum drying and electrode calendering. This work presents a new manufacturing method using a nonthermal plasma to create inter-particle binding without using any polymeric binding materials, enabling solvent-free manufacturing electrodes with any electrochemistry of choice. The cold-plasma-coating technique enables fabricating electrodes with thickness (>200 μm), high mass loading (>30 mg cm -2 ), high peel strength, and the ability to print lithium-ion batteries in an arbitrary geometry. This crosscutting, chemistry agnostic, platform technology would increase energy density, eliminate the use of solvents, vacuum drying, and calendering processes during production, and reduce manufacturing cost for current and future cell designs. Here, lithium iron phosphate and lithium cobalt oxide were used as examples to demonstrate the efficacy of the cold-plasma-coating technique. It is found that the mechanical peel strength of cold-plasma-coating-manufactured lithium iron phosphate is over an order of magnitude higher than that of slurry-casted lithium iron phosphate electrodes. Full cells assembled with a graphite anode and the cold-plasma-coating-lithium iron phosphate cathode offer highly reversible cycling performance with a capacity retention of 81.6% over 500 cycles. For the highly conductive cathode material lithium cobalt oxide, an areal capacity of 4.2 mAh cm -2 at 0.2 C is attained. We anticipate that this new, highly scalable manufacturing technique will redefine global lithium-ion battery manufacturing providing significantly reduced plant footprints and material costs.

25 ENERGY STORAGE↗
DOE OSTI2021

TRANSIENT THERMAL MODELING OF THE HIGH BURNUP DEMONSTRATION RESEARCH PROJECT CASK USING STAR-CCM+ AND COBRA-SFS

The Department of Energy in collaboration with the Electric Power Research Institute is in the process of conducting the High Burnup Demonstration Project. Where the objective is to characterize the performance of high-burnup fuel in long term storage. As part of this demonstration, a TN-32B dry storage cask was instrumented and loaded with spent nuclear fuel at North Anna Nuclear Generating Station in November 2017. The project cask provides a unique opportunity to gain information on spent fuel and cask performance from an in-service operational system. The cask was instrumented with thermocouples inside of the fuel assemblies, then loaded and dried using normal procedures. After the drying process a thermal soak period where the cask was left indoors was used to obtain steady temperatures for model comparison along with surface temperatures. This paper details thermal modeling validation work that was done to model both the steady state and transient cases. Two modeling tools were used to predict temperatures in the cask. The general purpose CFD and heat transfer code STAR-CCM+ was used with both a detailed pin-by-pin model was used along with a more efficient k-effective simplification. The other code is COBRA-SFS, a purpose-built detailed thermal modeling tool developed and maintained at PNNL. Results from all the modeling tools were compared blind to the test data. Each model utilized general design information and compared reasonably well to the blind steady state data using convection and surface temperature boundary conditions. Post-test adjustments were later made to better reflect the “as built” conditions of the cask. Based on lessons learned from the steady state analysis the vacuum drying process was modeled in both codes. Utilizing the measured loading conditions the transient models were able to compare very well with the measured data. Overall the project showed an ability to model spent fuel storage conditions very well and future work is planned to generalize the methodology used for vacuum drying.

Thermal Analysis, Spent Fuel, Dry Storage↗
NASA NTRS1980

Space Technology for Crop Drying

McDonnell Douglas came up with a new method of drying agricultural crops derived from vacuum chamber technology called MIVAC, a compression of microwave vacuum drying system. A distant cousin of the home microwave oven, MIVAC dries by means of electrically- generated microwaves introduced to a crop-containing vacuum chamber. Microwaves remove moisture quickly and the very low pressure atmosphere in the chamber permits effective drying at much lower than customary temperatures. Thus energy demand is doubly reduced by lower heat requirement and by the shorter time electric power is needed.

Source record↗
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 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↗
NASA NTRS2014

Cleaning Process Development for Metallic Additively Manufactured Parts

Additive Manufacturing of metallic components for aerospace applications offers many advantages over traditional manufacturing techniques. As a new technology, many aspects of its widespread utilization remain open to investigation. Among these are the cleaning processes that can be used for post finishing of parts and measurements to verify effectiveness of the cleaning processes. Many cleaning and drying processes and measurement methods that have been used for parts manufactured using conventional techniques are candidates that may be considered for cleaning and verification of additively manufactured parts. Among these are vapor degreasing, ultrasonic immersion and spray cleaning, followed by hot air drying, vacuum baking and solvent displacement drying. Differences in porosity, density, and surface finish of additively manufactured versus conventionally manufactured parts may introduce new considerations in the selection of cleaning and drying processes or the method used to verify their effectiveness. This presentation will review the relative strengths and weaknesses of different candidate cleaning and drying processes as they may apply to additively manufactured metal parts for aerospace applications. An ultrasonic cleaning technique for exploring the cleanability of parts will be presented along with an example using additively manufactured Inconel 718 test specimens to illustrate its use. The data analysis shows that this ultrasonic cleaning approach results in a well-behaved ultrasonic cleaning/extraction behavior. That is, it does not show signs of accelerated cavitation erosion of the base material, which was later confirmed by neutron imaging. In addition, the analysis indicated that complete cleaning would be achieved by ultrasonic immersion cleaning at approximately 5 minutes, which was verified by subsequent cleaning of additional parts.

Tramel, Terri L.↗
DOE OSTI2024

Quantification of Residual Water in Spent Fuel Dry Storage Canisters Using Mass Spectrometry

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, such as fuel degradation; cladding corrosion, embrittlement, or breaching; and the creation of a flammable environment via radiolysis of water. 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 details on the quantification of residual water in the Advanced Drying Cycle Simulator (ADCS), an apparatus 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. The ADCS is outfitted with thermocouples to measure the thermal response of the ADCS to simulated decay heats and internal helium fill pressures relevant to commercial drying procedures. The ADCS is also instrumented with pressure transducers to measure the pressures and vacuum levels observed during simulated commercial drying. The most unique instrumentation used for quantifying residual water in the ADCS is a Hiden Analytical HPR-30 mass spectrometer (MS), which measures gas compositions of the ADCS internal free volume, based on partial pressures calculated from relative proportions of gas molecules detected by the MS. This report details the methodology used to implement MS measurements in quantifying residual water in the ADCS. This methodology includes the calibration of the HPR-30 MS to a Buck Research Instruments CR-4 chilled mirror hygrometer, which itself is calibrated to a NIST-traceable standard. Data collected by both the MS and the chilled mirror hygrometer from water/helium mixtures ranging from 150 to 500,000 ppmv water in helium were used to generate calibration curves, establishing a source of verification of MS measured water contents. Details regarding water content measurement uncertainties are included in this report, defining the accuracy and verifiability of the HPR-30 MS in measuring residual water content in simulated dry storage canister environments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1974

Solid metabolic waste transport and stowage investigation

The basic Waste Collection System (WCS) design under consideration utilized air flow to separate the stool from the WCS user and to transport the fecal material to a slinger device for subsequent deposition on a storage bowel. The major parameters governing stool separation and transport were found to be the area of the air inlet orifices, the configuration of the air inlet orifice and the transport air flow. Separation force and transport velocity of the stool were studied. The developed inlet orifice configuration was found to be an effective design for providing fecal separation and transport. Simulated urine tests and female user tests in zero gravity established air flow rates between 0.08 and 0.25 cu sm/min (3 and 9 scfm) as satisfactory for entrapment, containment and transport of urine using an urinal. The investigation of air drying of fecal material as a substitute for vacuum drying in a WCS breadboard system showed that using baseline conditions anticipated for the shuttle cabin ambient atmosphere, flow rates of 0.14 cu sm/min (5 cfm) were adequate for drying and maintaining biological stability of the fecal material.

Burt, R. A.↗
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↗