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DOE OSTI2020

An Updated Analysis of Clad Degradation

This report provides an analysis of the clad barrier function associated with the direct disposal of dual purpose canisters (DPCs) under hypothetical conditions in a shale repository and in an alluvial repository, including the effect of a postulated criticality event inside a disposed DPC. Should a postulated criticality event occur in a hypothetical shale repository, cladding will primarily degrade by general corrosion. Stress corrosion cracking, hydride cracking, creep failure, pitting and crevice corrosion, rod pressurization, and clad unzipping are calculated to have little impact on cladding persistence. At the higher temperature expected during a postulated criticality event in a saturated shale repository, general corrosion of cladding would be rapid - on the order of 0.034 microns/yr. A few hundred years after onset of a postulated criticality event in a shale repository complete general corrosion of fuel assembly grid spacer walls and guide tubes will likely result in settling of fuel rods upon each other. This rod consolidation should displace the water moderator and possibly terminate a postulated criticality. The primary potential degradation pathway for cladding in a hypothetical alluvial repository is localized corrosion by fluoride, which cannot occur in a shale repository. Fluoride-enhanced corrosion of cladding would be accelerated under the slightly higher (< 100°C) temperatures associated with a postulated criticality event. The impact of criticality in both cases (shale and alluvial) would be to increase the amount of failed cladding. But it would require very specialized transport pathways.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Radiation Hardened Foam Cold Test Plan - Phase II: Foam Characterization Testing and Environmental Chamber Testing of FoamBag Fixative Foam

This document outlines the Phase II test objectives and implementation plan for a down-select foam fixative technology intended to facilitate activities in support of the Savannah River Site (SRS) F/H labs deactivation and decommissioning (D&D) efforts. It is a collaborative effort between Savannah River National Laboratory (SRNL), Florida International University (FIU), and the SRS F/H labs team intended to test and evaluate the potential of a polyurethane resin foam in mitigating the release of contamination during dismantling operations on radioactively contaminated piping in legacy facilities. The cold test plan addresses specific requirements highlighted by site and safety personnel and will be executed at FIU and SRNL test and lab facilities. Results from the cold tests will inform the hot test at F/H labs, which will use the foam fixative to confine and/or isolate residual contamination within a 3-dimensional void space of Hastelloy C-22 piping designated for removal from the area and transported to a designated disposal facility. Phase I testing was previously conducted using Hilti CP620 fixative foam. Results from Phase I testing indicate Hilti CP-620 fixative foam is incompatible with SRS site hot taps as an effective foam delivery method into Hastelloy C-22 piping. Phase II testing will be conducted using FoamBag TM fixative foam as an alternative fixative foam option. Phase II testing will address eight test objectives: (1) evaluation of the adhesion and bonding properties of FoamBag TM in Hastelloy C-22 piping, (2) evaluation of the adhesion of FoamBag TM in piping under various moisture conditions, (3) determination of the heat profile of FoamBag TM during curing, (4) determination of the internal pipe pressure after FoamBag TM deployment and curing, (5) conduct a leak test to determine if FoamBag TM is effective at creating a full seal within piping, (6) headspace testing of FoamBag TM to determine if there are any associated off-gas hazards during FoamBag TM curing, (7) conduct environmental chamber testing of FoamBag TM to understand how environmental parameters impacts FoamBag TM curing, and (8) conduct fire testing to evaluate FoamBag TM fire retardant characteristics.

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DOE OSTI2020

Monte Carlo characterization of the cosmic ray muon flux in shallow subsurface geological repositories intended for disposal of radioactive materials

Recent challenges in monitoring subsurface geological repositories intended for disposal of radioactive materials such as spent nuclear fuel call for new, innovative concepts that are facility independent, cost-effective, passive, and reliable. Once nuclear material is in place at these facilities, reverifying the inventory may no longer be feasible if continuity of knowledge is lost or unavailable to the inspectors. Using cosmic ray muons may present several potential advantages over conventional photon/neutron signatures, and their use in safeguards applications have only received attention in the past decade. However, there have been limited efforts to explore the integration of cosmic ray muons into repository safeguards and study potential gains, risks, and costs. This paper presents a Monte Carlo-based methodology to characterize the cosmic ray muon flux, including muon angular and energy differential distributions at depths representative of subsurface geological repositories. Since there have been limited measurements at these sites and a measurement made in one site is not always transferable to another site, the objective is to develop an efficient simulation method and useful parametrizations to provide a convenient tool for enabling muon simulations at any geological repository site. It is expected these results will provide a better understanding of how muons can be integrated into an existing geological repository safeguards framework.

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DOE OSTI2023

Fabrication and Testing of DOE Standard Canister Closure Leak Test Assembly – 24139

DOE manages over 300 types of spent nuclear fuel (SNF), many of which are located at the Idaho National Laboratory (INL) site. Managing this large variety of SNF for storage, transportation, and disposal poses a challenge to DOE. The Idaho Cleanup Project and INL are collaborating on the DOE SNF Road-Ready Demonstration (“Road-Ready Demonstration”), which will develop and demonstrate the designs, technology, processes, and regulatory framework for packaging DOE-managed SNF for “road-ready dry storage.” Road-ready dry storage is an SNF management concept in which SNF is packaged into dry, sealed canisters that are then placed in onsite storage in anticipation of later transport and disposition. The forward-looking goal of the Road-Ready Demonstration is to establish the foundation for a large-scale road-ready dry storage program at the INL site. In support of establishing a large-scale road-ready dry storage program at the INL site, the Road-Ready Demonstration will first package Fort St. Vrain SNF currently in dry storage at INL into several DOE Standard Canisters (DOESCs). These DOESCs will in turn be loaded into another containment similar to commercial multi-purpose canisters. This multi-purpose canister will then be compatible with a transportation or storage system, such as a storage cask for interim storage or transportation package for offsite transport. These DOESCs will remain sealed over the course of their storage, transportation, and applicable disposal functions. The closure process for the DOESC will include fuel and basket loading, welding, inspection, leak testing, and, if needed, repair. As a follow-up to previous discussions on the design of the DOE Closure Leak Test Assembly (LTA), this report describes recent fabrication and testing efforts performed at INL. DOESCs are sealed by two sequential gas tungsten arc welds, both of which are performed by remotely operated and semiautomatic welding systems. The first weld is a circumferential pipe weld that completes the assembly of the canister body and lid assembly. The second and final closure weld attaches the vent plug to the vent socket via a butt joint. After the second weld is performed, the welds are helium leak tested using an evacuated envelope technique. The LTA was designed for both remote and manual operation. This report describes the fabrication and performance testing associated with the evacuated envelope technique. INL staff designed, fabricated, and tested the LTA at INL facilities. This testing included establishing technique and system sensitivities in accordance with ASME and American National Standards Institute N14.5 requirements. Forthcoming work will cover such areas as design optimization, process and personnel qualification, and implementation in Road-Ready Demonstration operations.

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DOE OSTI2020

Nevada Higher Education Benefits from US Department of Energy (DOE) Environmental Management (EM) Nevada Program's Transfer of Geologic Samples - 20442

The Nevada National Security Site (NNSS), formerly the Nevada Test Site, was the location of 100 historic atmospheric and 828 historic underground nuclear tests from 1951 to 1992. Related to this historic nuclear testing, geologic and hydrologic studies of the site were conducted utilizing the skills and the expertise of the U.S. Geological Survey (USGS) and national laboratories (Los Alamos National Laboratory, Lawrence Livermore National Laboratory), who were principal in leading the development of weapons and were responsible for specific underground testing programs. The subsurface samples and technical/scientific data associated with NNSS geologic studies (past and present) are preserved and stored at the USGS Mercury Core Library and Data Center located at the NNSS. Currently, the facility stores over 2,000,000 linear feet of cores and cuttings from more than 2,600 drill holes that can be accessed for study. Most of the samples and historic geologic work was focused directly on subsurface geologic settings that relate directly to historic underground nuclear testing. Since 1992, the United States has observed a unilateral moratorium on full-scale nuclear testing, and the U.S. Department of Energy (DOE) Environmental Management (EM) Nevada Program is now responsible for hydrogeologic characterization of the potential impacts to the natural groundwater systems that may have resulted from historic underground nuclear tests. As part of this effort, the EM Nevada Program Underground Test Area (UGTA) Activity has drilled and completed over 58 deep (2,000 - 7,000 ft.) characterization wells, totaling in excess of 170,000 linear feet of cuttings and core samples. Cutting samples were collected as triplicate samples for each respective depth interval, to account for potential later nondestructive/destructive analysis and to preserve samples for regulatory purposes. Recently, it was recognized by the EM Nevada Program that opportunities may exist to reduce the cost and floor space required for the storage of geologic samples at the Mercury Core Library without impacting the integrity and representative nature of the samples necessary for project execution. An initiative was sponsored by EM Nevada Program to evaluate several options: 1) disposal of a portion of cuttings and cores in a land fill setting; 2) reduction in sample volume through skeletonizing core and cuttings from wells to a representative but much smaller number of samples; and 3) solicit potential interest in the academic community where other geoscientists could freely access the samples for studies. The Nevada state university system through the Nevada Bureau of Mines and Geology responded positively to the opportunity to receive these NNSS samples from the EM Nevada Program. In June 2019, approximately 17,000 geologic samples, representing greater than 170,000 linear feet of drilling and weighing over 20,000 pounds, were shipped from the NNSS to the Great Basin Science Sample and Records Library located in Reno, Nevada. The benefits resulting from this transfer were realized by the EM Nevada Program in terms of cost and space savings for geologic sample storage at the NNSS. The transfer did not impact the EM Nevada Program mission as access to representative geologic samples for regulatory and scientific purposes was preserved. Moreover, the Nevada Bureau of Mines and Geology acquired a significant resource of subsurface geologic samples and supporting technical data to support academic and scientific studies in a complex volcanic setting in southwest Nevada. (authors)

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DOE OSTI2025

Nitrated Anion Exchange Resin Disposal Plan Calcination Studies

Reillex HPQ resin was nitrated in a manner matching procedures in use by the Aqueous Nitrate Recovery Facility. The resin was kept wet and then calcined to determine if nitrate was present in the residue. The potential for dispersion during heating and generation of soot or ash was evaluated. A series of experiments were performed by calcining the resin to 650 °C, and both residual solids and liquids were collected and measured for nitrates by Raman spectroscopy and test strip measurements.

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DOE OSTI2020

Overview of the Long-Term Stewardship of the Remediated Uranium Mining and Milling Sites in Europe - 20079

Uranium mining in Europe began in the ' 40's for civil and military applications. The great majority of the uranium mines have now been closed and are going through the stages of decommissioning and/or rehabilitation. Today, the main uranium producers are outside the EU; only a few mines are still in operation inside the EU. In 2006, the situation of the uranium mines and mill tailings in Europe, identifying the tailings liabilities, their status, the plans for these sites and any associated hazards, was reported to the EC. The legacy of uranium mines and mill tailing disposals in the EU are located mainly in 12 Member States where, in total, 87 tailing objects were identified at 63 sites. The volume of each one of these tailings disposal varies from a few thousand to several tens of million m{sup 3}. The majority of these sites are no longer in operation. More than 40 tailing embankments have already been completely remediated. Approximately 20 are currently being remediated while for the remaining objects, remediation has yet to start. After ten years, and for a few cases where the respective projects have not yet been completed, it seems that most of Europe's uranium mines and milling legacies were remediated according to acceptable environmental standards. Nevertheless, Europe is about to face a new challenge that comes from the concluded remediation projects for which an effective set of measures for coordinated institutional control, known as 'long-term stewardship', is needed including, in most cases, continuous and active maintenance. Moreover, the uncertainty of what will happen to these sites if institutional control fails at some point in time is a major concern. This work presents a summary of the situation and current issues of the uranium mine and mill tailings in Europe and their remediation status. It addresses not only the management of existing uranium mine and mill legacies but also the methods and strategies that have been used to prevent the creation of new future legacy sites from uranium mines still in operation. (authors)

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DOE OSTI2024

Drop Analysis of Department of Energy Standard Canister with Fort Saint Vrain SNF

DOE manages over 300 types of SNF, most of which are located at the INL site. The Idaho Cleanup Project and INL are collaborating on the Road-Ready Capability Demonstration Project, which will develop and demonstrate the designs, technology, processes, and regulatory framework for packaging DOE-managed SNF at the INL site for ?road-ready dry storage?. ?Road-ready dry storage? is a SNF management concept where SNF is packaged into dry and sealed canisters, which are then placed in on-site storage in anticipation of later transportation. The forward-looking goal of the Demonstration is establishing the foundation for a large-scale road-ready dry storage program at the INL site. The Demonstration will first package Fort Saint Vrain SNF currently stored at INL into several DOE Standard Canisters. These Standard Canisters will then be loaded into another commercial transportation or storage containment system (e.g., storage cask or transportation package). The Standard Canister is a class of standardized canisters designed for containing the large variety of DOE-managed SNF during interim storage, transportation and/or disposal at a geological repository. One critical aspect of road-ready dry storage is the ability to license the DOE Standard Canisters and its associated transportation package to 10 CFR 71. Depending on the SNF and transportation strategy, the Standard Canisters may have to maintain structural integrity under normal conditions of transport and hypothetical accident scenarios (i.e., drop events). The Standard Canisters have been tested and analyzed under various SNF loading configurations and accident drop events in support of the Idaho Spent Fuel Facility and other DOE programs. However, no analysis has been completed to support the recent Demonstration. This analysis will consider the Ø0.5 m × 5.1 m Standard Canister under drop scenario(s) considered in previous INL tests and analyses, including the 9 m drop at 80 degree off vertical. However, this analysis will consider the more recent Fort Saint Vrain loading configurations proposed for the Demonstration. This analysis will performed using strain-based acceptance criteria established by the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section III, Division 3. It will be compared to previous analyses and form the foundation of further formal calculations that will be used to support licensing efforts of the road-ready dry storage system at INL.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Dual-Purpose Canister Filling Demonstration Project Progress Report

This report discusses the progress made at the Oak Ridge National Laboratory to support direct disposal of dual-purpose canisters (DPCs). Acceptable filler materials should demonstrate that the probability of criticality in DPCs during the disposal timeframe is below the probability threshold for inclusion in a repository performance assessment. This effort, which will ultimately result in a full-scale demonstration, includes computational fluid dynamics (CFD) modelling developed to gauge the filling process and to uncover any unforeseen issues. Filling simulations of the lower region (mouse holes) of a prototypic DPC show successfully simulate filling of the void space inside the canister and a smooth, even progression of the liquid level. Flow through a pipe that is similar to the drainpipe in a DPC is being investigated separately to gain valuable insight of the flow regime inside a pipe. Three physical experiments validating the computational filling model have been completed using surrogate liquids. One experiment using molten metal has been completed and the results demonstrate adequate filling of the canister void spaces. Although the scale experiment observed some void spaces related to shrinkage of the metal during cooling, the volume filled is expected to be sufficient to meet the purpose of moderator exclusion. Further experiments will scale up the geometry, moving towards a full-sized DPC demonstration, and provide a more rigorous investigation of candidate filler materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Middlesex South FUSRAP Site: Collaboration Towards Beneficial Reuse - 20338

The US Department of Energy (DOE) Office of Legacy Management (LM) manages DoE's post-closure responsibilities and ensures the future protection of human health and the environment with respect to sites that have no continuing DOE mission after undergoing remediation. LM's beneficial reuse program promotes the LM strategic goal to sustainably manage and optimize use of public lands. The beneficial reuse program aims to repurpose former contaminated sites to restore the environment, protect the public health, revitalize communities, and spur economic growth. Benefits from the reuse of a site may increase the local tax base, facilitate job growth, utilize existing infrastructure, and enhance or protect natural resources. Beneficial reuse also promotes protectiveness by ensuring activities are compatible with long-term maintenance and protection of public health and the environment, as well as by retaining good stewardship of natural resources. LM actively participates in promoting the economic development vision of the surrounding communities by collaborating with local communities to promote regional or municipal initiatives. The Borough of Middlesex, New Jersey, has prepared the Lincoln Boulevard Redevelopment Plan, which will reinvent a once-viable downtown and enhance the quality of life for the community. The Middlesex South, New Jersey, Formerly Utilized Sites Remedial Action Program (FUSRAP) site lies within the boundaries of this redevelopment area. In support of this development plan, the Borough is interested in acquiring the site for its Department of Public Works and to increase street access in the future commercial district to be built near adjacent properties. The opportunity to put the FUSRAP site into productive use aligns with LM's mission and strategic goals. In support of beneficial reuse activities, DOE can dispose of excess real property using several mechanisms. The two disposal options applicable to the Middlesex south site include (1) utilizing the US General Services Administration (GSA) or (2) using the Title 10 Code of Federal Regulations Section 770, 'Transfer of Real Property at Defense Nuclear Facilities for Economic Development' (known as a '770 transfer') process. Using the 770 transfer requires economic development as the primary driver, and a specific redevelopment proposal must be submitted. If there is no interest in specific economic development, the GSA option becomes the default mechanism for disposal of the property. In the case of the Middlesex south site, the 770 transfer process is being utilized and the Borough has submitted its economic development proposal to LM. Although LM is encouraged by the opportunity to put the site into productive use, the Middlesex south site is currently on the US Environmental Protection Agency (EPA) National Priorities List and remediation is being performed by the US Army Corp of Engineers (USACE) under the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA). Consequently, any property transfer must meet the requirements under CERCLA Section 120(h), 'Property Transferred by Federal Agencies.' As the groundwater remedy will not be in place for several years, approval would be required from the US EPA administrator and the governor of the state of New Jersey to perform an early transfer of the site (before all response actions have been completed). The time frame for these approvals to occur under the normal process conflicts with the proposed redevelopment schedule; therefore, various options for expedited sale, transfer, or lease of portions of the site to the Borough were evaluated. The subsequent decision on a mutually beneficial path forward was the result of a collaborative effort between LM, USACE, the Borough and its redevelopment team, EPA, and the New Jersey Department of Environmental Protection. (authors)

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DOE OSTI2020

The Relationship Between Dose Rate and Decay Heat for Spent Nuclear Fuel Casks

Decay heat and dose rate are two important limits used for determining the allowable contents of spent fuel (SNF) in dry storage systems and transportation packages. While the decay heat limit is used to maintain fuel cladding integrity and ensure retrievability, dose rates are used to demonstrate compliance with regulatory requirements on radiation protection. Because both dose rate and decay heat result from decay of radioisotopes in SNF, this study is an attempt to examine the relationship between dose rate and decay heat for a given cask design. Dose rates were evaluated for 198 cask configurations, that include various SNF system designs (e.g., storage, transfer, transport), SNF characteristics (e.g., fuel types, burnup, cooling time), and loading maps (e.g., uniform loading, zone loading), while a constant decay heat was maintained. The decay heat was calculated using US Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 3.54, Revision 2, and verified using ORIGEN sequence within SCALE code system. The ORIGEN outputs were used as source terms in the dose analysis using 198 different configurations. A computer script was developed to calculate the cooling time necessary to achieve a given decay heat for a given enrichment, assembly average burnup, assembly mass, and in-core history using a rootfinder algorithm. Initially the computer script was developed to provide cooling time and burnup calculations directly to the analysis of dose rates and decay heats, so a comparison between Used Nuclear Fuel-Storage Transportation and Disposal Analysis Resource Data System (UNF-ST&DARDS) results and RG3.54 data was made; results are included in the appendix to this document. However, an iterative approach was used to compute cooling time, and its accuracy did not depend on the results of the RG3.54r2 algorithm, although the algorithm was still used. The results of the evaluation presented herein clearly demonstrate that a given decay heat does not correspond to a unique dose rate for a variety of cask and package designs. There is no clear pattern to develop a correlation between decay heat and the source terms. Depending on burnup, enrichment, cask type, and loading pattern, dose rates varied for the exact same decay heat—in some cases by 400% for a given cask. For cases in which decay heat was held constant through selection of the appropriate cooling time, dose rates would decrease with increasing burnup, and in other cases, dose rates would increase. The large variation in dose rates for a constant decay heat indicates that casks loaded based on decay heat—that is allowing any burnup, cooling time, and enrichment combinations that yield the qualified decay heat limit(s) —cannot ensure that an Independent Spent Fuel Storage Installation or a spent fuel transportation package will meet the regulatory limits set forth by the respective regulations, i.e., 10 CFR 72 or 10 CFR 71.

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DOE OSTI2022

Impacts of Crystalline Host Rock on Repository Barrier Materials at 250 °C: Hydrothermal Co-Alteration of Wyoming Bentonite and Steel in the Presence of Grimsel Granodiorite

Direct disposal of dual-purpose canisters (DPC) has been proposed to streamline the disposal of spent nuclear fuel. However, there are scenarios where direct disposal of DPCs may result in temperatures in excess of the specified upper temperature limits for some engineered barrier system (EBS) materials, which may cause alteration within EBS materials dependent on local conditions such as host rock composition, chemistry of the saturating groundwaters, and interactions between barrier materials themselves. Here we report the results of hydrothermal experiments reacting EBS materials—bentonite buffer and steel—with an analogue crystalline host rock and groundwater at 250 °C. Experiment series explored the effect of reaction time on the final products and the effects of the mineral and fluid reactants on different steel types. Post-mortem X-ray diffraction, electron microprobe, and scanning electron microscopy analyses showed characteristic alteration of both bentonite and steel, including the formation of secondary zeolite and calcium silicate hydrate minerals within the bentonite matrix and the formation of iron-bearing clays and metal oxides at the steel surfaces. Swelling clays in the bentonite matrix were not quantitatively altered to non-swelling clay species by the hydrothermal conditions. The combined results of the solution chemistry over time and post-mortem mineralogy suggest that EBS alteration is more sensitive to initial groundwater chemistry than the presence of host rock, where limited potassium concentration in the solution prohibits conversion of the smectite minerals in the bentonite matrix to non-swelling clay species.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2020

235-F GoldSim Fate and Transport Model: Uncertainty Quantification

Building 235-F was configured with two missions in mind: Actinide Billet Line (ABL) and the fabrication of Pu-238 oxide for space program applications. ABL produced Np-237 billets for use in SRS reactors, whereas the design process, fabrication, and examination of Pu-238 oxide powder occurred within the following areas, respectively: Plutonium Experimental Facility (PEF), Plutonium Fuel Form (PuFF), and Old Metallography Lab (OML). By 1990 production ceased and by 2006 de-inventory occurred; however, assays have shown significant holdup remains within ABL and PuFF. As a result, 235-F is a Category 2 nuclear facility, with plans to undergo deactivation and decommission (D and D) via In-Situ Disposal (ISD). The purpose of this project is to ensure United States Environmental Protection Agency (USEPA) groundwater radiation maximum contaminant level (MCL) and dosage standards are met during the D and D of 235-F by quantifying uncertainty through probabilistic modeling and evaluation of various ISD alternatives. GoldSim is a dynamic modeling software package with a graphical, object-oriented interface capable of capturing the influence of complex system input variability on probabilistic system outcomes. A GoldSim stochastic fate and transport model for 235-F was developed and matched with a PORFLOW deterministic model to simulate probabilistic release and flow of radionuclides from ABL and PuFF into the vadose zone, the Upper Three Runs (UTR) Aquifer, and UTR Creek. The GoldSim model was used to probabilistically evaluate four ISD scenarios against USEPA groundwater radiation MCLs and dosage standards. The deterministic 235-F GoldSim fate and transport model continues to be refined to match the results of the PORFLOW deterministic model to ensure the model accurately represents radionuclide movement through the groundwater system. The stochastic variables that are utilized within the GoldSim model are founded on the most current data; a conservative perspective is taken where needed. Alignment with the PORFLOW deterministic model, coupled with input stochastic variability, allows the probabilistic 235-F GoldSim model to capture the conservative breadth of possible outcomes for radionuclide fate within this particular system. This ensures that the USEPA MCLs and dosage limits hold even in the worst case scenarios.

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DOE OSTI2022

Preconceptual Design of Irradiated Fuel Salt Management System

The National Reactor Innovation Center was established by the U.S. Department of Energy to accelerate the demonstration and deployment of advanced nuclear reactors. To meet this mission, the National Reactor Innovation Center has developed the Laboratory for Operations and Testing in the U.S. (LOTUS) to support the first fast-spectrum molten salt reactor demonstration. The goal of this research is to understand the system requirements that may be applied to a molten salt reactor experiment in the LOTUS test bed and to perform systems analyses and preconceptual development of a storage container for the management of the irradiated molten salt fuel. This study proposes a set of irradiated fuel salt management processes for the molten salt reactor, from defueling to storage or disposal. For that, the historic Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory has served as a crucial source of data for a mock application. This paper proposes a potential geometry for storing irradiated fuel salt based on MSRE data and the Monte Carlo N-Particle radiation transport code. The geometry was analyzed for criticality safety under abnormal situations, such as water ingress and interaction between containers. Based on the determined configuration of the storage system, the decay heat produced from the irradiated fuel salts was calculated to provide a reasonable time scale of the dormancy period. Using the same geometry assumed above, the neutron and gamma dose rates from the irradiated fuel salt at the end of the dormancy period were estimated, which will be useful for understanding the handling and radiation protection requirements for handling the irradiated fuel salt. Lastly, a reactor in the LOTUS test bed could use chloride-based fuel as opposed to the fluoride-based fuel used in MSRE. Thus, a literature study provided data to understand the similarities and differences between the characteristics of fluoride- and chloride-based salts in terms of salt chemistry, material characteristics, postirradiation behavior, neutronics, thermal hydraulics, and heat removal. This study, together with the proposed irradiated fuel salt management processes, preliminary criticality calculations, decay heat, and dose rate estimates, provides a strong basis for future investigation into the management of irradiated fuel salts generated during molten salt reactor technology demonstrations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Influences of Mg Ions and Al Ions on Deposition Behavior of Supersaturated Silicic Acid - 20261

Silicic acid can change into various species such as deposit, colloidal, and soluble form under certain conditions: for example, gradual pH changes around geological disposal repositories. In particular, silicic acid deposits narrow groundwater flow paths, thereby restricting the flow of groundwater. Such effects may contribute to a delay in radionuclide migration. Previous studies have reported that silicic acid supersaturated in sodium and calcium ions can narrow the flow paths. This study examined the effect of magnesium (Mg) and aluminum (Al) ions on the deposition of supersaturated silicic acid on the solid phase. The deposition behavior of supersaturated silicic acid on the solid phase was investigated in a batch system while changing the initial supersaturation concentration, the coexisting ion concentration, and the amount of added solid phase. In the presence of Mg ions, silicic acid precipitation increased with increasing Mg ion concentration due to the influence of the electrolyte, and crosslinking reactions. The coexistence of Al ions significantly suppressed silicic acid deposition due to the formation of a covering on the solid surface during Al ion hydrolysis under conditions of high Al concentration (1.4 mM). On the basis of these experimental results, the apparent deposition rate constant of silicic acid was estimated to be in the range from 10{sup -11} to 10{sup -10} m/s. The Damkoehler number (the dimensionless number expressed by the ratio of the characteristic time of mass transfer and the chemical reaction constant) was sufficient to cause the clogging of the host rock by deposition. This suggests that supersaturated silicic acid deposition, regardless of the presence of Mg ions (and Al ions), contributes to improving barrier performance against radionuclide migration in the host rock. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Effect of chromium on corrosion resistance of Ni-Cr-Mo-Gd alloys in seawater

Neutron absorbing materials are being considered within commercial spent nuclear fuel disposal canisters to maintain nuclear subcriticality in storage. To select candidate alloys for the canisters, both neutron absorption and corrosion resistance should be considered. This work examines corrosion resistance of Ni-Cr-Mo-Gd alloys developed specifically for neutron absorption. The addition of Gd results in a secondary gadolinide phase (Ni 5 Gd) that significantly changes the corrosion properties. Testing was performed primarily in seawater at 30°C. Seawater was selected as the most prevalent terrestrial brine and is characterized by a high chloride concentration. Various electrochemical corrosion techniques were carried out to evaluate Ni-Cr-Mo-Gd alloys with different Cr compositions and investigate the role of Ni 5 Gd phase on corrosion behavior. C22 was included as a benchmark material, due to the similarity in composition and the significant corrosion data available. Here, test results showed a tendency to passivate over time which is attributed to dissolution of surface exposed Ni 5 Gd phase. Cross-sectional analysis indicated that dissolution could penetrate hundreds of micrometers deep under aggressive conditions. It was found that higher Cr variant (21.01%) showed much shallower impact, suggesting Cr prevented primary phase corrosion and thus reduced Ni 5 Gd phase dissolution. Acid pickling of the specimens showed much less dissolution for a higher Cr material and suggested some primary phase dissolution for the low Cr specimen. Acid pickled specimens showed positive shifts in the repassivation potential, suggesting increased surface passivation.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2026

Hybrid-Energy-Powered Electrochemical Ocean Alkalinity Enhancement Model: Plant Operation, Cost, and Profitability

Electrochemical ocean alkalinity enhancement is a form of marine carbon dioxide removal, a rapidly growing industry that is powered by efficient onshore or offshore energy sources. As more and larger deployments are being planned, it is important to consider how variable energy sources like tidal energy can impact plant performance and costs. An open-source Python-based generalizable model for electrodialysis-based ocean alkalinity enhancement has been developed that can capture key system-level insights of the electrochemistry, ocean chemistry, acid disposal, and co-product creation of these plants under various conditions. The model additionally accounts for hybrid energy system performance profiles and costs via the National Laboratory of the Rockies’ H2Integrate tool. The model was used to analyze an example theoretical plant deployment in North Admiralty Inlet, including how the plant is impacted by the available energy sources in the region and the scale at which plant costs are covered by the co-products it generates, such as recycled concrete aggregates, without requiring carbon credits. The results show that the example plant could be profitable without carbon credits at commercial scales of 100,000 to 1 million tons of carbon dioxide removal per year, so long as it uses low-cost electricity sources and either sells acid or recovers recycled concrete aggregates with about 1 molar acid concentrations, though more research is needed to confirm these results.

hybrid energy↗