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DOE OSTI2024

Sister Rod Destructive Examinations (FY23) Appendix I: SNF Aerosols Released During Rod Fracture

This report documents work performed under the Spent Fuel and Waste Disposition’s Spent Fuel and Waste Science and Technology program for the US Department of Energy (DOE) Office of Nuclear Energy (NE). This work was performed to fulfill Level 3 Milestone M3SF-23OR010201026, “FY23 M3 draft report on results from testing in FY23,” within work package SF-23OR01020102 and is an updated to the work reported in M2SF-23OR010201024, M2SF-22OR010201047, M2SF-21OR010201032, M2SF-19ORO010201026, and M2SF-19OR010201028. As a part of the DOE NE High Burnup Spent Fuel Data Project, Oak Ridge National Laboratory (ORNL) is performing destructive examinations (DEs) of high burnup (HBU) (>45 GWd/MTU) spent nuclear fuel (SNF) rods from the North Anna Nuclear Power Station operated by Dominion Energy. The SNF rods, called sister rods or sibling rods, are all HBU and include four different kinds of fuel rod cladding: standard Zircaloy-4 (Zirc-4), low-tin Zirc-4, ZIRLO, and M5. The DEs are being conducted to obtain a baseline of the HBU rods’ condition before dry storage and are focused on understanding overall SNF rod strength and durability. Fuel rods and defueled cladding will be tested to derive material properties. Although the data generated can be used for multiple purposes, one primary goal for obtaining the post-irradiation examination data and the associated measured mechanical properties is to support SNF dry storage licensing and relicensing activities by (1) addressing identified knowledge gaps and (2) enhancing the technical basis for post-storage transportation, handling, and subsequent disposition.This report documents the status of the ORNL Phase 1 DE activities related to the collection of SNF aerosol particles released during fuel rod fracture in 4-point bending in Phase 1 of the sister rod test program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Continued Investigations of Respirable Release Fractions for Stress Corrosion Crack-Like Geometries

The formation of a stress corrosion crack (SCC) in the canister wall of a dry cask storage system (DCSS) has been identified as a potential issue for the long-term storage of spent nuclear fuel. The presence of an SCC in a storage system could represent a through-wall flow path from the canister interior to the environment. Modern, vertical DCSSs are of particular interest due to the commercial practice of using relatively high backfill pressures (up to approximately 800 kPa) in the canister to enhance internal natural convection. This pressure differential offers a comparatively high driving potential for blowdown of any particulates that might be present in the canister. In this study, the rates of gas flow and aerosol transmission of a spent fuel surrogate through an engineered microchannel with dimensions representative of an SCC were evaluated experimentally using coupled mass flow and aerosol analyzers. The microchannel was formed by mating two gage blocks with a linearly tapering slot orifice nominally 13 μm (0.005 in.) tall on the upstream side and 25 μm (0.0010 in.) tall on the downstream side. The orifice is 12.7 mm (0.500 in.) wide by 8.89 mm (0.350 in.) long (flow length). Surrogate aerosols of cerium oxide, CeO 2 , were seeded and mixed with either helium or air inside a pressurized tank. The aerosol characteristics were measured immediately upstream and downstream of the simulated SCC at elevated and ambient pressures, respectively. These data sets are intended to demonstrate a new capability to characterize SCCs under well-controlled boundary conditions. Modeling efforts were also initiated that evaluate the depletion of aerosols in a commercial dry storage canister. These preliminary modeling and ongoing testing efforts are focused on understanding the evolution in both size and quantity of a hypothetical release of aerosolized spent fuel particles from failed fuel to the canister interior and ultimately through an SCC.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Annual Performance Testing of Tracer Gas Detectors for use in Radionuclide NESHAP Compliance Testing

Los Alamos National Laboratory’s (LANL’s) compliance with Radionuclide NESHAP0F 1 regulations is managed by the Radioactive Air Emissions Management (RAEM) team, part of LANL’s Compliance Programs group (EPC-CP). One area of the Radionuclide NESHAP addresses requirements for siting a stack sample system. Prior to commissioning a new stack sampling system, the ANSI Standard for stack sampling requires that the stack sample location must meet several criteria, including uniform mixing of tracer gas (sulfur hexafluoride, SF 6 ) and tracer aerosol (liquid oil droplets) in the air stream. This memorandum will specifically focus on the tracer gas portion of the ANSI Standard. For these mix tests, tracer gas is injected into the stack air stream and the resulting air concentrations are measured across the plane of the stack at the proposed sampling location. The coefficient of variation of the media concentration must be under 20% when evaluated over the central 2/3 area of the stack or duct.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2013

HLRW management during MR reactor decommissioning in NRC 'Kurchatov Institute'

A program of decommissioning of MR research reactor in the Kurchatov institute started in 2008. The decommissioning work presumed a preliminary stage, which included: removal of spent fuel from near reactor storage; removal of spent fuel assemble of metal liquid loop channel from a core; identification, sorting and disposal of radioactive objects from gateway of the reactor; identification, sorting and disposal of radioactive objects from cells of HLRW storage of the Kurchatov institute for radwaste creating form the decommissioning of MR. All these works were performed by a remote controlled means with use of a remote identification methods of high radioactive objects. A distribution of activity along high radiated objects was measured by a collimated radiometer installed on the robot Brokk-90, a gamma image of the object was registered by gamma-visor. Spectrum of gamma radiation was measured by a gamma locator and semiconductor detector system. For identification of a presence of uranium isotopes in the HLRW a technique, based on the registration of characteristic radiation of U, was developed. For fragmentation of high radiated objects was used a cold cutting technique and dust suppression system was applied for reduction of volume activity of aerosols in air. The management of HLRW was performed by remote controlled robots Brokk-180 and Brokk-330. They executed sorting, cutting and parking of high radiated part of contaminated equipment. The use of these techniques allowed to reduce individual and collective doses of personal performed the decommissioning. The average individual dose of the personnel was 1,9 mSv/year in 2011, and the collective dose is estimated by 0,0605 man x Sv/year. Use of the remote control machines enables reducing the number of working personal (20 men) and doses. X-ray spectrometric methods enable determination of a presence of the U in high radiated objects and special cans and separation of them for further spent fuel inspection. The sorting of radwaste enabled shipping of the LLRW and ILRW to special repositories and keeping of the HLRW for decay in the Kurchatov institute repository. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Examining Nuisance Aerosol Detections in Light of the Origin of the Screening Process (November 2021)

The evolution of philosophy and computations in the International Data Center (IDC) related to aerosol samples have had profound impacts on the number of recorded detections in the network since routine operations began in 2000. Key decisions from policymakers have been the list of triggering radionuclides, the scheme for categorizing these into interest levels 1-5, and an algorithm for determining when an anthropogenic isotope is seen so often that it is no longer interesting, known as the Exponential Weighted Moving Average (EWMA). These are described in the Operations Manual of the IDC. Key parameters that are controlled by the IDC but for which the IDC receives occasional input from policymakers include the constants in EWMA and the peak significance threshold for individual gamma rays, the latter of which directly leads to determination of the presence or absence of a radionuclide in a sample. There are also changes in computations which the IDC makes and informs policy makers about, such as changes in how background is computed, which could also affect the ease of detecting a peak – real or false. Rather than focus on the quantitative changes due to computation changes, this work records some thinking on how isotopes and peak significance levels were chosen, and the resulting detections seen over 18 years during the buildup of the International Monitoring System (IMS). These detections are considered on a global scale to try to determine the relative impact on monitoring, and in some cases, the nature of their existence. Repeated detections of 131I and 133I are the most troublesome, but they are not so frequent to be a major problem for the Verification Regime. These detections could probably be handled adequately using scientific methods currently under development for xenon backgrounds. It is also somewhat problematic that top-level analysis of aerosol backgrounds has not been reported previous to this. The steep increase in the rate of detections after 2016 are a concern, either in the actual backgrounds or from changes in the calculations methods used to generate the Reviewed Radionuclide Report (RRR.) Final conclusions of the authors are that the computational stability of the RRR is very important. With computational stability, changes can be usefully analyzed as being due to changes in radioactivity in Earth’s atmosphere This report is a distillation into text of a talk given in the Radionuclide Experts Group (RNEG) in Vienna during Working Group B (WGB) in February of 2019. This report does not directly contain any IDC data, only summaries by year, or by isotope, or by location. No specific IDC detection by time, location, or isotope is included.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

Aerosol Particle Deposition on a Spent Nuclear Fuel Assembly Spacer Grid

The flow and particle deposition patterns on surfaces in an idealized spacer grid for a 17x17 pressurized water reactor (PWR) assembly in a spent fuel canister are modeled using computational fluid dynamics (CFD) with laminar flow. The effects of gravitational settling, non-Stokesian flow, and particle slip are first rigorously analyzed. From the analysis, non-Stokesian effects and slip may be neglected for the particle sizes and conditions expected in a canister. For particles that do not settle out, a swirling flow pattern at the corners of a spacer grid channel directs particles to the leeward side of the flow vanes where much of the deposition occurs. Particle deposition increases with increasing particle diameter. Deposition also increases with decreasing flow velocity as this provides more time for particles to settle and deposit on the leeward side of the flow vanes. The fraction of particles that are transmitted through a spacer grid is determined as a function of inlet gas velocity and particle diameter by running the CFD calculation for each set of conditions and for each particle diameter. Curve fits of the transmission curve as a function of particle diameter for a specified spacer grid and flow velocity are applied to a lognormal particle mass density function for the inlet particles. The resulting mass density function and aerosol mass fraction that passes through the spacer grid can be determined analytically without resorting to numerical iteration. A sample calculation of the analytical solution is demonstrated for a lognormal particle mass density function.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Pipe Overpack Container Fire Testing: Phase I II & III.

The Pipe Overpack Container (POC) was developed at Rocky Flats to transport plutonium residues with higher levels of plutonium than standard transuranic (TRU) waste to the Waste Isolation Pilot Plant (WIPP) for disposal. In 1996 Sandia National Laboratories (SNL) conducted a series of tests to determine the degree of protection POCs provided during storage accident events. One of these tests exposed four of the POCs to a 30-minute engulfing pool fire, resulting in one of the 7A drum overpacks generating sufficient internal pressure to pop off its lid and expose the top of the pipe container (PC) to the fire environment. The initial contents of the POCs were inert materials, which would not generate large internal pressure within the PC if heated. POCs are now being used to store combustible TRU waste at Department of Energy (DOE) sites. At the request of DOE’s Office of Environmental Management (EM) and National Nuclear Security Administration (NNSA), starting in 2015 SNL conducted a series of fire tests to examine whether PCs with combustibles would reach a temperature that would result in (1) decomposition of inner contents and (2) subsequent generation of sufficient gas to cause the PC to over-pressurize and release its inner content. Tests conducted during 2015 and 2016 were done in three phases. The goal of the first phase was to see if the PC would reach high enough temperatures to decompose typical combustible materials inside the PC. The goal of the second test phase was to determine under what heating loads (i.e., incident heat fluxes) the 7A drum lid pops off from the POC drum. The goal of the third phase was to see if surrogate aerosol gets released from the PC when the drum lid is off. This report will describe the various tests conducted in phase I, II, and III, present preliminary results from these tests, and discuss implications for the POCs.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

A consideration of radionuclide particulate resuspension as a verification tool in the CTBT on-site inspection verification component

Here, a considerable amount of radioactivity could leak from an underground nuclear test at levels easily detectable by one or more of the radiological methods available for on-site inspection, even if the event does not immediately result in gas or particulate debris reaching stations of the International Monitoring System, of the Comprehensive Nuclear-Test-Ban Treaty. This work presents a feasibility study showing that winds or human activity can resuspend surface debris in quantities that could be detected using high-volume aerosol samplers outside an established restricted area and subsequent measurement in a Base of Operations laboratory.

CTBT↗
DOE OSTI2014

Design and Integrate Improved Systems for Nuclear Facility Ventilation and Exhaust Operations

Objective: The objective of this R&D project would complete the development of three new systems and integrate them into a single experimental effort. However, each of the three systems has stand-alone applicability across the DOE complex. At US DOE nuclear facilities, indoor air is filtered and ventilated for human occupancy, and exhaust air to the outdoor environment must be regulated and monitored. At least three technical standards address these functions, and the Los Alamos National Laboratory would complete an experimental facility to answer at least three questions: (1) Can the drag coefficient of a new Los Alamos air mixer be reduced for better operation in nuclear facility exhaust stacks? (2) Is it possible to verify the accuracy of a new dilution method for HEPA filter test facilities? (3) Is there a performance-based air flow metric (volumetric flow or mass flow) for operating HEPA filters? In summary, the three new systems are: a mixer, a diluter and a performance-based metric, respectively. The results of this project would be applicable to at least four technical standards: ANSI N13.1 Sampling and Monitoring Releases of Airborne Radioactive Substances from the Stacks and Ducts of Nuclear Facilities; ASTM F1471 Standard Test Method for Air Cleaning Performance of a High-Efficiency Particulate Air Filter System, ASME N511: In-Service Testing of Nuclear Air Treatment, Heating, Ventilating, and Air-Conditioning Systems, and ASME AG-1: Code On Nuclear Air And Gas Treatment. All of the three proposed new systems must be combined into a single experimental device (i.e. to develop a new function of the Los Alamos aerosol wind tunnel). Technical Approach: The Radiation Protection RP-SVS group at Los Alamos has an aerosol wind tunnel that was originally (2006) designed to evaluate small air samplers (cf. US EPA 40 CFR 53.42). In 2009, the tunnel was modified for exhaust stack verifications per the ANSI N13.1 standard. In 2010, modifications were started on the wind tunnel for testing HEPA filters (cf. ASTM F1471 and ASME N511). This project involves three systems that were developed for testing the 24*24*11 (inch) HEPA filters (i.e. the already mentioned mixer, diluter and metric). Prototypes of the mixer and the diluter have been built and individually tested on a preliminary basis. However, the third system (the HEPA metric method) has not been tested, since that requires complete operability of the aerosol wind tunnel device. (The experimental wind tunnel has test aerosol injection, control and measurement capabilities, and can be heated for temperature dependent measurements.) Benefits: US DOE facilities that use HEPA filters and/or require exhaust stacks from their nuclear facility buildings will benefit from access to the new hardware (mixer and diluter) and performance-based metric (for HEPA filter air flow).

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2023

Progress Report on Model Development for Aerosol Transport through Divergent Cracks Paths

This report summarizes the progress in developing a phenomenological model of aerosol transport, deposition, and plugging through microchannels. The purpose of this effort is to introduce to a user community—involving researchers, regulators, and industry—a generic, reliable numerical model for the prediction of aerosol transport while accounting for potential deposition and plugging of the leak paths to model spent nuclear fuel (SNF) release from postulated Stress Corrosion Cracks in canisters under storage or transportation. Current work focuses on expanding the model to predict aerosol (and gas) flow through complex microchannel (nozzle) geometries in addition to the rectangular and cylindrical geometries, as we approach more realistic Stress Corrosion Crack conditions. In this regard, a divergent nozzle geometry, used at Sandia National Laboratories (SNL) for testing aerosol release and retention (Durbin et al. 2021), was added to the model. The model was then validated with blowdown data for the particular microchannel, from experiments conducted at SNL. The report also presents preliminary non-benchmarked aerosol penetration fraction and mass flow rate for a monodisperse 10-micron (AED) particle concentration of 1.7e-08 kg/m 3 being released through the divergent microchannel geometry from the SNL’s aerosol experimental tank setup.Future work will involve validating the aerosol model and updating the model’s Graphical User Interface (GUI) to include the divergent nozzle geometry. This is expected to help stakeholders perform quick and easy first principles calculations without needing to understand the underlying MATLAB script.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2010

INITIAL RESULTS: AN ULTRA-LOW-BACKGROUND GERMANIUM CRYSTAL ARRAY

ABSTRACT Treaty verification techniques, environmental surveillance, and physics experiments continue to require increased sensitivity for detecting and quantifying radionuclides of interest. This can be accomplished with new detector designs that establish high detection efficiency and reduced instrument backgrounds. Current research is producing an intrinsic germanium (HPGe) array designed for high detection efficiency, ultra-low-background performance, and sensitive gamma-gamma coincidence detection. The system design is optimized to accommodate filter paper samples, e.g., samples collected by the Radionuclide Aerosol Sampler/Analyzer (RASA). The system will provide high sensitivity for weak collections on atmospheric filter samples (e.g., < 105 fissions), as well as offering the potential to gather additional information from higher-activity filters using gamma cascade coincidence detection. The first of two HPGe crystal arrays in ultra-low-background vacuum cryostats has been assembled, with the second in progress. Traditional methods for constructing ultra-low-background detectors were followed, including use of materials known to be low in radioactive contaminants, use of ultra-pure reagents, and use of a clean room assembly. The cryostat is constructed mainly from copper electroformed into near-final geometry at Pacific Northwest National Laboratory. Details of the detector assembly, vacuum and thermal performance testing, and initial measurement results are presented.

Keillor, Martin E.↗
DOE OSTI2017

Analysis of an MCU HEPA filter

A series of direct analyses on three portions (inlet, center, and outlet) of the High Efficiency Particulate Air (HEPA) filter material from the Modular Caustic-Side Solvent Extraction Unit (MCU) have been performed; this includes x-ray methods such as X-Ray Diffraction (XRD), Contained Scanning Electron Microscopy (CSEM) and X-Ray Fluorescence (XRF), as well as Fourier Transform InfraRed spectroscopy (FTIR). Additionally, two leaching studies (one with water, one with dichloromethane) have been performed on three portions (inlet, center, and outlet) of the HEPA filter material, with the leachates being analyzed by Inductively-coupled plasma emission spectroscopy (ICPES), Semi-Volatile Organic Analysis (SVOA) and gammascan. From the results of the analyses, SRNL feels that cesium-depleted solvent is being introduced into the HEPA filter. The most likely avenue for this is mechanical aerosolization of solvent, where the aerosol is then carried along an airstream into the HEPA filter. Once introduced into the HEPA filter media, the solvent wicks throughout the material, and migrates towards the outlet end. Once on the outlet end, continual drying could cause particulate flakes to exit the filter and travel farther down the airstream path.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Results from Evaluation of Representative ASME AG-1 Section FK Radial Flow Dimple Pleated HEPA Filters Under Elevated Conditions - 12002

The American Society of Mechanical Engineers (ASME) has recently added Section FK establishing requirements for radial flow HEPA filters to the Code on Nuclear Air and Gas Treatment (AG-1). Section FK filters are expected to be a major element in the HEPA filtration systems across the US Department of Energy (DOE) complex. Radial flow filters have been used in Europe for some time, however a limited amount of performance evaluation data exists with respect to these new AG-1 Section FK units. In consultation with a technical working group, the Institute for Clean Energy Technology (ICET) at Mississippi State University (MSU)has evaluated a series of representative AG-1 Section FK dimple pleated radial flow HEPA filters. The effects of elevated relative humidity and temperature conditions on these filters are particularly concerning. Results from the evaluation of Section FK filters under ambient conditions have been presented at the 2011 waste management conference. Additions to the previous test stand to enable high temperature and high humidity testing, a review of the equipment used, the steps taken to characterize the new additions, and the filter test results are presented in this study. Test filters were evaluated at a volumetric flow rate of 56.6 m{sup 3}/min (2000 cfm) and were challenged under ambient conditions with Alumina, Al(OH){sub 3}, until reaching a differential pressure of 1 kPa (4 in. w.c.), at which time the filters were tested, unchallenged with aerosol, at 54 deg. C (130 deg. F) for approximately 1 hour. At the end of that hour water was sprayed near the heat source to maximize vaporization exposing the filter to an elevated relative humidity up to 95%. Collected data include differential pressure, temperature, relative humidity, and volumetric flow rate versus time. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

Small Column Ion Exchange Design and Safety Strategy

Small Column Ion Exchange (SCIX) is a transformational technology originally developed by the Department of Energy (DOE) Environmental Management (EM-30) office and is now being deployed at the Savannah River Site (SRS) to significantly increase overall salt processing capacity and accelerate the Liquid Waste System life-cycle. The process combines strontium and actinide removal using Monosodium Titanate (MST), Rotary Microfiltration, and cesium removal using Crystalline Silicotitanate (CST, specifically UOP IONSIV{reg_sign}IE-911 ion exchanger) to create a low level waste stream to be disposed in grout and a high level waste stream to be vitrified. The process also includes preparation of the streams for disposal, e.g., grinding of the loaded CST material. These waste processing components are technically mature and flowsheet integration studies are being performed including glass formulations studies, application specific thermal modeling, and mixing studies. The deployment program includes design and fabrication of the Rotary Microfilter (RMF) assembly, ion-exchange columns (IXCs), and grinder module, utilizing an integrated system safety design approach. The design concept is to install the process inside an existing waste tank, Tank 41H. The process consists of a feed pump with a set of four RMFs, two IXCs, a media grinder, three Submersible Mixer Pumps (SMPs), and all supporting infrastructure including media receipt and preparation facilities. The design addresses MST mixing to achieve the required strontium and actinide removal and to prevent future retrieval problems. CST achieves very high cesium loadings (up to 1,100 curies per gallon (Ci/gal) bed volume). The design addresses the hazards associated with this material including heat management (in column and in-tank), as detailed in the thermal modeling. The CST must be size reduced for compatibility with downstream processes. The design addresses material transport into and out of the grinder and includes provisions for equipment maintenance including remote handling. The design includes a robust set of nuclear safety controls compliant with DOE Standard (STD)-1189, Integration of Safety into the Design Process. The controls cover explosions, spills, boiling, aerosolization, and criticality. Natural Phenomena Hazards (NPH) including seismic event, tornado/high wind, and wildland fire are considered. In addition, the SCIX process equipment was evaluated for impact to existing facility safety equipment including the waste tank itself. SCIX is an innovative program which leverages DOE's technology development capabilities to provide a basis for a successful field deployment.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Surface Sampling Techniques for the Canister Deposition Field Demonstration

This report describes plans for dust sampling and analysis for the multi-year Canister Deposition Field Demonstration. The demonstration will use three commercial 32PTH2 NUHOMS welded stainless steel storage canisters, which will be stored at an ISFSI site in Advanced Horizontal Storage Modules. One canister will be unheated; the other two will have heaters to achieve canister surface temperatures that match, to the degree possible, spent nuclear fuel (SNF) loaded canisters with heat loads of 10 kW and 40 kW. Surface sampling campaigns will take place on a yearly or bi-yearly basis. The goal of the planned dust sampling and analysis is to determine important environmental parameters that impact the potential occurrence of stress corrosion cracking on SNF dry storage canisters. Specifically, the size, morphology, and composition of the deposited dust and salt particles will be quantified, as well as the soluble salt load per unit area and the rate of deposition, as a function of canister surface temperature, location, time, and orientation. Sampling locations on the canister surface will nominally include 25 locations, corresponding to 5 circumferential locations at each of the 5 longitudinal locations. At each sampling location, a 2x2 sampling grid (containing 4 sample cells) will be painted onto the metal surface. During each sampling campaign, two samples at each sampling location will be collected, in a specific routine to measure both periodic (yearly or bi-yearly) and cumulative deposition rates. For each sample, a wet and a dry sample will be collected. Wet samples will be analyzed to determine the composition of the soluble salt fraction and to estimate salt loading per unit area. Dry samples will be analyzed to assess particle size, morphology, mineralogy, and identity (e.g. for floral/faunal fragments). The data generated by this proposed sampling plan will provide detailed information on dust and salt aerosol deposits on spent nuclear fuel canister surfaces. The anticipated results include information regarding particle compositions, size distributions, and morphologies, in addition to particle deposition rates as a function of canister surface location, orientation, time, and temperature. The information gathered during the Canister Deposition Field Demonstration is critical for ongoing efforts to develop a detailed understanding of the potential for stress corrosion cracking on SNF dry storage canisters

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

Estimation of Respirable Aerosol Release Fractions through Stress Corrosion Crack-Like Geometries

The formation of a stress corrosion crack (SCC) in the canister wall of a dry cask storage system (DCSS) has been identified as a potential issue for the long-term storage of spent nuclear fuel. The presence of an SCC in a storage system could represent a through-wall flow path from the canister interior to the environment. Modern, vertical DCSSs are of particular interest due to the significant backfill pressurization of the canister, up to approximately 800 kPa. This pressure differential offers a relatively high driving potential for blowdown of any particulates that might be present in the canister. In this study, the carrier gas flow rates and aerosol transmission properties were evaluated for an engineered microchannel with characteristic dimensions similar to those of an SCC. The microchannel was formed by mating two gage blocks with a slot orifice measuring 28.9 μm (0.0011 in.) tall by 12.7 mm (0.500 in.) wide by 8.86 mm (0.349 in.) long (flow length). Surrogate aerosols of cerium oxide, Ce0 2 , were seeded and mixed inside a pressurized tank. The aerosol characteristics were measured immediately upstream and downstream of the simulated SCC at elevated and ambient pressures, respectively. These data sets are intended to demonstrate a new capability to characterize SCCs under well-controlled boundary conditions. Separate modeling efforts are also underway that will be validated using these data. The test apparatus and procedures developed in this study can be easily modified for the evaluation of more complex SCC-like geometries including laboratory-grown SCC samples.

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

Identification of Potential Waste Processing and Waste Form Options for Molten Salt Reactors

The overall summary of the waste management envelope discussed in this report is represented by the diagram shown in Figure S1. The streams that are included in this discussion include the following from reactor systems in which chlorine-based salts with and without processing and fluorine-based salts with and without processing are being considered: 1) Off-gas streams (e.g., particulates, aerosols, reactive gases, 3 H, water, residual halides, N 2 , O 2 , noble gases); 2) Salt waste streams (e.g., chloride- or fluoride-based); 3) Separated salt streams (e.g., 7 Li recycle, 37 Cl recycle, eutectic salts, fission products); 4) Metal waste streams (e.g., reactor components, plenums); 5) Carbon waste streams (e.g., graphite, SiC); 6) Decommissioning and decontaminating (D&D) waste streams; 7) Operating waste streams

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

A Phenomenological Model for the Transport and Plugging of Aerosol through Stress Corrosion Cracks

Recent experimental evidence from Sandia (2018) demonstrated partial microchannel plugging due to particulate deposition. A direct consequence is that the accuracy of consequence assessments can be improved by accounting for the leak path deposition of aerosol in the source term. The filtration effect of microchannels, such as cracks, is relevant to the source term assessments. An important side effect of aerosol deposition in leak paths could be the plugging of the leak path. This report summarizes current progress (as of end of FY19) on the development of a phenomenological model of aerosol transport, deposition, and plugging through microchannels. The purpose is to introduce a generic, reliable numerical model for prediction of aerosol transport, deposition, and plugging in leak paths that are similar to stress corrosion cracks while accounting for potential plugging formation. This effort includes (i) development of a numerical model to analyze the various deposition processes in leak paths and provide quantitative estimates of penetration factors, as well as an understanding of the variables that affect them, (ii) confirmation of model validity with theoretical and experimental data, and (iii) parametric analysis for different scenarios including various particle sizes, pressures, and crack dimensions. Such a model could be used to improve the accuracy of consequence assessments and reduce the uncertainty of radiological consequence predictive analyses by taking the filtering effect of leak path aerosol deposition and plugging into account in the source term. This report includes: (i) an overview of main stress corrosion cracking characteristics, (ii) an overview of the developed numerical model to analyze the various deposition processes in leak paths, to provide quantitative estimates of penetration factors, and to gain an understanding of the variables that affect them, (iii) a summary of validation results and verification of the model’s validity with recent experimental results from Sandia and others, and (iv) a model application example.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗