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At least 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

ATF-1 Power Histories

MCNP5 calculated ATF-1 power histories are documented in this Technical Evaluation (TEV). The current revision of this TEV (Revision 0), includes power histories through Advanced Test Reactor (ATR) Cycle 162A. The Cycle 162A power histories are based off projection calculations while all other cycles are based off as-run (core follow) simulations. This TEV will serve as the single source for ATF-1 experiment power history information and will be updated after each cycle of operation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Fracture Mechanics Analysis of Isotope Fuel Impact Tester, Milli-Watt, Multi-Hundred Watt and GPHS Outer Catch-Tube Subassembly Design

The Isotope Fuel Impact Tester (IFIT) at PF-4 is a specialized set of equipment designed for simulating high-speed impact of experimental parts. The IFIT is an inert gas launcher designed to impact 238 PuO 2 heat source assemblies, fuel encapsulation materials, structural materials, and subassemblies of radioisotope thermoelectric generators (RTGs). TA-55 management desires to re-use outer catch tube (OCT) of the milli-watt (MW), multi-hundred watt (MHW) and general purpose heat source (GPHS) designs and thus requires a technical evaluation of potential life cycles to failure. This report provides a fitness-for-service (FFS) assessment using fracture mechanics techniques based on the state-of-stress developed with the aid of a computational finite element model (FEM) of the IFIT. Importantly, the report describes the different material options for the OCT; H-11 vacuum-arc re-melted (VAR) and H-13 VAR for a given heat treatment resulting in a Rockwell-C hardness of 42, and an additional option of 4340 alloy steel with a range of heat treatment resulting in Rockwell-C hardness of 40 to 44.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

TEV-3776 Hybrid – ZIRCEX Process Demonstration Siting Study

The purpose of this technical evaluation (TEV) is to detail the siting considerations for supporting pre-conceptual design of the engineering-scale Hybrid – ZIRCEX Process (formerly known as Hybrid Uranium Recovery Demonstration [HURD]) demonstration. This process represents the first-of a kind integration of a select group of technologies meant to recover high enriched uranium (HEU) from zirconium and aluminum clad used nuclear fuel (UNF) for down-blending to produce high assay low enriched uranium (HALEU).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Initial Considerations for Large-Scale Carbon Removal in the United States

The Biden Administration has established a goal of removing 1 Gt carbon dioxide equivalents per year (CO 2e /yr) from the atmosphere and of achieving net zero greenhouse gas emissions (GHG) by 2050. Our team of leading academic and DOE national laboratory experts is conducting the first economy-wide technical evaluation of the options for achieving this carbon dioxide removal (CDR) goal. The net zero goal includes targets of 100% clean electricity by 2035 with 40% GHG emissions reductions by 2030, which helps set primary scenario boundaries for our analysis of the CDR supply curve. Our analysis will evaluate feasibility, performance, and costs on a county level for the entire USA (including AK and HI where possible), considering all removal methods that are currently well-enough developed for us to estimate the likely costs in 2050. We anticipate that more than 1 Gt CO 2e of removal will be available to the Nation. We will identify how much of each CO 2 removal approach is available in specific regions of the Nation and provide cumulative costs and volumes (a supply curve) by region for 2050. We expect to complete this detailed analysis by late 2023.

42 ENGINEERING↗
DOE OSTI2011

Sensitivity of HTGR Heat and Power Production to Reactor Outlet Temperature, Economic Analysis

This technical evaluation (TEV) has been prepared as part of a study for the Next Generation Nuclear Plant (NGNP) Project to evaluate the economics of integrating a high-temperature gas-cooled reactor (HTGR) with conventional chemical processes. This TEV addresses the economics of heat and power produced using an HTGR, as well as the effect of increasing the reactor outlet temperature (ROT) on the economic results. These results are preliminary and should be refined as the design of the HTGR progresses, if the design of the HTGR is changed significantly, or if additional refinements of the HTGR capital and/or operating costs become available. The HTGR capital, operating and maintenance (O&M) costs, fuel, and decommissioning costs are based on the correlations and costs presented for a nth of a kind HTGR in TEV-1196 (Idaho National Laboratory [INL] 2011a). The production of heat and power, as well as the effect of increasing the ROT on process results, has previously been addressed in detail in TEV-981 (INL 2011b). In that report, detailed process models for heat and power production using an HTGR were developed, with a range of reactor outlet temperatures from 650 to 950°C, in 50°C increments. This report is a follow-up to TEV-981 and evaluates the economics of heat and power production, employing either a Rankine or Brayton cycle, for ROTs of 700 to 950°C, in 50°C increments for a single 600 MWt HTGR and a four-pack of 600 MWt HTGRs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Multi-Scale Integrated Monitoring System for Enhancing Methane Emission Detection, Quantification & Prediction

This report details the progress and findings of a comprehensive study on reviewing existing solutions, identifying technology gaps, and formulating an “all-in-one” integrated strategy for developing the next-generation multiscale methane monitoring and modeling platform, conducted under grant number DE-FE0032292. Co-led by Dr. David Ebert, Dr. Binbin Weng, and Dr. Chenghao Wang at the University of Oklahoma, the project’s goal was to develop an integrated approach for building this engineering platform to detect, quantify, and mitigate methane emissions across various temporal scale, spatial scales, and sectors. The planning grant study began with an extensive review of various methane sensing and monitoring technologies and systems, surveying over 100 technology providers globally. This review revealed the prevalence of optical methods over chemical methods in commercially available sensors, with Non-Dispersive Infrared (NDIR), Tunable Diode Laser Absorption Spectroscopy (TDLAS), and Optical Gas Imaging (OGI) cameras being the most prevalent options. A trend towards more advanced optical techniques was observed, driven by increased regulatory focus and technological advancements. The technical evaluation of these sensing technologies provided crucial insights into their capabilities and limitations. The study examined emerging technologies such as Differential Absorption LiDAR (DIAL), which show promise for high-precision and long-range detection. The team then investigated the features and application bandwidth of various sensing platforms, including handheld, fixed/stationary, mobile, aerials, and spaceborne monitors. Pilot field studies were conducted to assess the capabilities of solutions for different emission scenarios. Field work with sensor deployments was conducted at three distinct site types: an oil & gas industry site, a cattle ranching operation, and a waste processing facility. The team also conducted a thorough review of methane flux inverse modeling approaches, focused on physically based methods. These approaches were categorized into simple, intermediate, and advanced methods. A realtime WRF-GHG (Weather Research and Forecasting-Greenhouse Gas) modeling system was developed and applied, incorporating multiple data sources to guide field experiments and inform methane plume detection. The project identified and analyzed numerous categories of methane data sources, including satellite measurements, ground-based sensors, and inventory databases. Key platforms examined include EDGAR, EPA GHGI, NASA TROPOMI, Carbon Mapper, and Climate TRACE, among others. The team proposed an architecture for a comprehensive methane monitoring platform. This system incorporates multi-source data acquisition, advanced data processing and assimilation, interactive visualization tools, and analytical capabilities for emissions forecasting and scenario analysis. The proposed platform aims to provide a user-friendly interface catering to various stakeholders, from researchers to policymakers. The architecture includes sophisticated data ingestion methods, a centralized data warehouse, and advanced analytical tools for data fusion and interpretation. To ensure the relevance and effectiveness of the proposed system, a comprehensive survey was conducted to gather stakeholder input on system requirements. Key findings include a strong need for integrating various data types and formats, a preference for real-time data updates and advanced visualization tools, and a demand for user-friendly interfaces catering to different expertise levels.

03 NATURAL GAS↗
DOE OSTI2022

Roadmap to Iodine and Mercury Abatement Materials Selection in Nuclear Waste Processing Off-Gas Streams

This work provides a guide for candidate mercury (Hg) and iodine-129 (129I) abatement material identification, screening, evaluation, technical gap identification, and bench-scale testing prioritization while also conceptualizing a materials deployment roadmap for implementation of new materials in the Hanford Waste Treatment and Immobilization Plant (WTP) Low-Activity Waste (LAW) Facility at Hanford in Richland, Washington, and elsewhere in the U.S. Department of Energy (DOE) complex for similar applications. The study was prompted by the need to replace the Kombisorb BAT-37 due to uncertainties in Hg and 129I capture performance and future availability for use in the WTP LAW Facility secondary off-gas system, specifically the Carbon Adsorber units. However, replacement of this material could also mitigate two other issues with Kombisorb BAT-37 (and its successor BAT II 37): (1) fire safety risk due to exothermic heat generated by adsorption reactions between the carbon material and Hg and (2) the risk of low retention of 129I in glass and subsequent downstream impacts on secondary liquid waste treatment at the Effluent Treatment Facility.

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

Assessment of Technologies for Improving Light-Duty Vehicle Fuel Economy—2025-2035

From daily commutes to cross-country road trips, millions of light-duty vehicles are on the road every day. The transportation sector is one of the United States’ largest sources of greenhouse gas emissions, and fuel is an important cost for drivers. The period from 2025-2035 could bring the most fundamental transformation in the 100-plus year history of the automobile. Battery electric vehicle costs are likely to fall and reach parity with internal combustion engine vehicles. New generations of fuel cell vehicles will be produced. Connected and automated vehicle technologies will become more common, including likely deployment of some fully automated vehicles. These new categories of vehicles will for the first time assume a major portion of new vehicle sales, while internal combustion engine vehicles with improved powertrain, design, and aerodynamics will continue to be an important part of new vehicle sales and fuel economy improvement.This study is a technical evaluation of the potential for internal combustion engine, hybrid, battery electric, fuel cell, nonpowertrain, and connected and automated vehicle technologies to contribute to efficiency in 2025-2035. In addition to making findings and recommendations related to technology cost and capabilities, Assessment of Technologies for Improving Light-Duty Vehicle Fuel Economy - 2025-2035 considers the impacts of changes in consumer behavior and regulatory regimes.

MARCHANT, GARY↗
DOE OSTI2022

Qualification of Continuous Fiber Reinforced 3D Printed Material for use in Hot Cell Environments

The ability to employ additively manufactured polymers in highly radioactive environments facilitates rapid prototyping as well as emergency replacement of various equipment components contained in-cell. A technical evaluation (TEV) exists seeking to qualify 3D printed polymers and carbon reinforcements for use in-cell, the data generated in this project seeks to support the TEV. The goal of the study is to determine if the loss in mechanical properties is low enough to justify semi-permanent use in-cell.

3D printing↗
DOE OSTI2013

Transition from Consultation to Monitoring-NRC's Increasingly Focused Review of Factors Important to F-Area Tank Farm Facility Performance - 13153

In consultation with the NRC, DOE issued a waste determination for the F-Area Tank Farm (FTF) facility in March 2012. The FTF consists of 22 underground tanks, each 2.8 to 4.9 million liters in capacity, used to store liquid high-level waste generated as a result of spent fuel reprocessing. The waste determination concluded stabilized waste residuals and associated tanks and auxiliary components at the time of closure are not high-level and can be disposed of as LLW. Prior to issuance of the final waste determination, during the consultation phase, NRC staff reviewed and provided comments on DOE's revision 0 and revision 1 FTF PAs that supported the waste determination and produced a technical evaluation report documenting the results of its multi-year review in October 2011. Following issuance of the waste determination, NRC began to monitor DOE disposal actions to assess compliance with the performance objectives in 10 CFR Part 61, Subpart C. To facilitate its monitoring responsibilities, NRC developed a plan to monitor DOE disposal actions. NRC staff was challenged in developing a focused monitoring plan to ensure limited resources are spent in the most cost-effective manner practical. To address this challenge, NRC prioritized monitoring areas and factors in terms of risk significance and timing. This prioritization was informed by NRC staff's review of DOE's PA documentation, independent probabilistic modeling conducted by NRC staff, and NRC-sponsored research conducted by the Center for Nuclear Waste Regulatory Analyses in San Antonio, TX. (authors)

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

NRC Waste Incidental to Reprocessing Program: Overview of Consultation and Monitoring Activities at the Idaho National Laboratory and the Savannah River Site - What We Have Learned - 12470

In 2005 the U.S. Nuclear Regulatory Commission (NRC) began to implement a new set of responsibilities under the Ronald W. Reagan National Defense Authorization Act (NDAA) of Fiscal Year 2005. Section 3116 of the NDAA requires the U.S. Department of Energy (DOE) to consult with the NRC for certain non-high level waste determinations and also requires NRC to monitor DOE's disposal actions related to those determinations. In Fiscal Year 2005, the NRC staff began consulting with DOE and completed reviews of draft waste determinations for salt waste at the Savannah River Site. In 2006, a second review was completed on tank waste residuals including sodium-bearing waste at the Idaho Nuclear Technology and Engineering Center Tank Farm at the Idaho National Laboratory. Monitoring Plans were developed for these activities and the NRC is actively monitoring disposal actions at both sites. NRC is currently in consultation with DOE on the F-Area Tank Farm closure and anticipates entering consultation on the H-Area Tank Farm at the Savannah River Site. This paper presents, from the NRC perspective, an overview of how the consultation and monitoring process has evolved since its conception in 2005. It addresses changes in methods and procedures used to collect and develop information used by the NRC in developing the technical evaluation report and monitoring plan under consultation and the implementation the plan under monitoring. It will address lessons learned and best practices developed throughout the process. The NDAA has presented significant challenges for the NRC and DOE. Past and current successes demonstrate that the NDAA can achieve its intended goal of facilitating tank closure at DOE legacy defense waste sites. The NRC believes many of the challenges in performing the WD reviews have been identified and addressed. Lessons learned have been collected and documented throughout the review process. Future success will be contingent on each agencies commitment to consistently apply the lessons learned and continue to create an open and collaborative work environment to maintain the process of continuous improvement. (authors)

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

Waste Treatment and Immobilization Plant, a Leader in the International Glass Community - 16003

The U.S. Department of Energy (DOE), Office of River Protection (ORP) is constructing the Tank Waste Treatment and Immobilization Plant (WTP) [1] to treat radioactive waste currently stored in underground tanks at the Hanford Site in Washington State. The WTP that is being designed to separate the tank waste into high-level waste (HLW) and low-activity waste (LAW) fractions with the majority of the mass (approximately 90 percent) directed to LAW and most of the activity (greater than 95 percent) directed to HLW. The pretreatment process, envisioned in the baseline, involves the dissolution of aluminum bearing solids to allow the aluminum salts to be processed through the cesium ion exchange and report to the LAW Facility. There is an oxidative leaching process to affect a similar outcome for chromium-bearing wastes. These two unit operations were advanced to accommodate shortcomings in glass formulation for HLW inventories. A significant consideration to the most recent revision of the glass models is the extensive collaboration developed, funded, managed, and coordinated from the ORP. The current effort includes researchers at the Pacific Northwest National Laboratory, Idaho National Laboratory, Savannah River National Laboratory, Brookhaven National Laboratory, Catholic University of America, Washington State University, Rutgers University, Institute of Chemical Technology Prague, University of Ottawa, Tokyo Institute of Technology, The University of Adelaide, and University of Sheffield. Independent technical evaluation having been solicited from faculty of Alfred University and Vanderbilt University. These advances in glass chemistry generate a need to revisit the feed vectors, flowsheet, and the unit operations driven by minimal waste loading requirements set forth in the contract for the design and construction of the plant. The current status and plans for the immediate future of glass chemistry are the subject of the primary concern to successfully completing our clean-up mission at Hanford. (authors)

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

EPRI Project Update on International Nuclear Power Plant Waste Classification - 16224

Countries around the globe use varying schemes for the characterization of radioactive waste. Many countries use adaptations of the International Atomic Energy Agency (IAEA) waste classification scheme. Other countries use the Class A, B, and C waste classification scheme that was developed in the United States. General comparisons have been made between the IAEA and the United States (U.S.) classification schemes, but a clear and well researched comparison of the two schemes was not evident in the available research data sets. The Electric Power Research Institute (EPRI) Radiation Safety (RS) Program has conducted a technical evaluation that compares these two waste characterization schemes. EPRI developed a set of reference wastes that are representative of nuclear power plant low-level radioactive waste in form and radioactivity concentrations (e.g. dry active waste, high activity resin, low activity resin, high activity filters and low activity filters). These reference wastes were then classified using the waste classification systems in place in several countries around the world. This paper and presentation will provide a summary of that research, show where strong comparisons and correlations can be made, and provide a discussion on how this information might inform the harmonization of global waste classification standards. (authors)

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

WIPP PA Sensitivity Studies in Support of CRA-2014 Completeness Determination and Compliance Decision - 18209

The Waste Isolation Pilot Plant (WIPP) has been developed by the U.S. Department of Energy (DOE) for the geologic (deep underground) disposal of transuranic (TRU) waste. Containment of TRU waste at the WIPP is regulated by the U.S. Environmental Protection Agency (EPA). The DOE demonstrates compliance with the containment requirements by means of performance assessment (PA) calculations. WIPP PA calculations estimate the probability and consequence of potential radionuclide releases from the repository to the accessible environment for a regulatory period of 10,000 years after facility closure. Since TRU waste was first received at the WIPP in 1999, compliance recertification applications (CRAs) have been required by the EPA on five-year cycle. The latest CRA (CRA-2014) was submitted by the DOE to the EPA in 2014. As part of the application, a formal set of PA calculations (CRA-2014 PA) was performed and results were documented in detail. The CRA-2014 PA incorporated several parameter updates based on new data, as well as the implementation of a different model for panel closures, barriers planned for the opening of each waste panel. The results presented as part of the CRA-2014 showed via the required complementary cumulative distribution function (CCDF) release curves that the estimated releases were within compliance limits. As part of their process for reviewing the completeness of the CRA-2014 [1], the EPA formally requested that the DOE perform sensitivity studies based on EPA-requested changes to model parameters. The DOE performed and documented the results of three sensitivity studies described below: CRA14-SEN2 [2], CRA14-SEN3 [3], and CRA14-SEN4 [4]. In each case, the new analysis results were compared to CRA- 2014 PA (CRA14) results to determine the impact of the EPA-requested changes. CRA14-SEN2: the application of EPA-requested modified parameters in the operations and experimental (non-waste) areas of the repository to simulate an accelerated (instantaneous) creep closure, the inclusion of capillary pressure effects on relative permeability, and an increase in initial/residual brine saturation and residual gas saturation. The modifications to the repository model resulted in increased pressures and decreased brine saturations in waste areas and increased pressures and brine saturations in the operations and experimental areas. However, the overall effects on total high-probability mean releases from the repository were entirely insignificant. It was concluded that the modeling assumptions associated with the operations and experimental areas of the repository have an insignificant effect on the prediction of total releases from the repository and/or adequacy of the current (CRA14) model to demonstrate compliance with the regulatory limits. CRA14-SEN3: the application of EPA-requested modified parameters in the panel closure areas and operations and experimental (non-waste) areas of the repository to simulate an accelerated (instantaneous) creep closure, the inclusion of capillary pressure effects on relative permeability, and an increase in initial/residual brine saturation and residual gas saturation. The modifications to the repository model predominantly result in increased pressures and decreased brine saturations in waste areas, increased pressures and brine saturations in the operations and experimental areas, and decreased brine and gas flows across all panel closures. The pressure increases in repository waste regions yielded slightly decreased brine saturations (on average) in those areas. However, the overall effects on total high-probability mean releases from the repository are entirely insignificant. It was concluded that the modeling assumptions made for this sensitivity study which were associated with the panel closures and north end areas of the repository have only a small effect on the prediction of total releases from the repository and/or adequacy of the current (CRA14) model to demonstrate compliance with the regulatory limits. CRA14-SEN4: the application of several changes to the current WIPP PA model (CRA14), including: parameter changes, use of an updated code, and a revised computational grid for the BRAGFLO code. The modifications to the repository model resulted in increased releases in all primary release mechanisms. The impacts of each EPA-requested change to CRA14 were analyzed with respect to each release mechanism in the CRA14-SEN4 study. Overall, total high-probability predicted mean releases from the repository were increased by about 15%. Total low-probability predicted mean releases were increased by about 107%. It was concluded that the EPA-requested changes to the CRA14 result in increases to the predicted total releases from the repository. However, releases calculated in the CRA14-SEN4 analysis remained below regulatory limits, demonstrating continued compliance of the WIPP. In January 2017, the EPA determined that CRA-2014 was complete and then began a technical evaluation of CRA-2014 to determine whether the WIPP continued to comply with the radiation protection standards for disposal. During their evaluation, the EPA utilized the sensitivity study results to inform their ultimate determination in July 2017 that the WIPP continues to comply with regulatory standards. (authors)

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

Experience by BAM in the Transport Package Safety Assessment of German Waste Containers - 18469

Transport and storage containers for low and mid-level radioactive waste are getting more and more of relevance due to the nuclear phase out decision in 2011. For higher activities Type B(U) approved waste containers will be needed for the shut down and dismantling phases of NPPs. It is expected that large quantities are required in the near future. German waste containers are generally approved for transport and interim storage and are also intended for the final disposal in the Konrad repository. BAM is involved in the authority licensing of transport package designs in Germany. In this context, BAM is responsible for the assessment of safe containment, mechanical, thermal and quality management issues. BAM also operates test facilities and performs drop as well as thermal tests during package licensing procedures. This paper summarizes our experience in the transport package design assessment of Type B(U) waste containers. A general overview of the approval process, the requirements and approaches BAM applies are described. Some examples are used to illustrate different aspects and technical issues we are addressing during the package assessment. In the first part of the paper the specific design aspects of German waste containers are described. Here, a general overview is given e.g. about closure systems, impact limiter designs, and the handling concepts. Furthermore, the wide range of radioactive content and their physical behavior including the impact on the packages assessment are described. The second part is focused on questions about the licensing and assessment process of German waste containers from the BAM authority point of view. The general approaches for the strategy of demonstration are outlined on the basis of the test conditions according to IAEA Regulations SSR-6. Furthermore, particular issues of the mechanical and thermal assessment with respect to the specific test conditions are discussed. For accident conditions of transport, aspects to be mentioned are e.g. the assessment of the lid bolts, the axial gap applied between content and lid, and the thermal behavior of the wood filled impact limiter after the fire test. However, issues of the assessment for routine and normal conditions of transport are addressed in this paper, too. The intention of this paper is introduce recent approval procedure experience in Germany, describing technical evaluation issues and so reduce rounds of questions during applications. (authors)

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

Developing and Testing a New Gas Distribution System for the Fluidized Bed of the Integrated Waste Treatment Unit (IWTU) Steam Reformer - 19219

The Integrated Waste Treatment Unit (IWTU) has been tasked with using steam reforming technology to treat 3.4 million liters of radioactive sodium-bearing waste (SBW) at the INL. Progress in developing steam reforming for SBW treatment has been hampered by numerous processing issues associated with the existing process equipment during simulant testing. The primary issue was high differential temperatures that developed in the fluidized bed of the denitration mineralization reformer (DMR). Based on technical evaluations and computer modeling, it was determined that the primary cause of poor fluidization in the DMR was the existing fluidized gas distributor. This distributor consisted of a combination of a ring header and four removable fluidizing gas rails. The combined ring header/rail system was unable to effectively fluidize the bed at the bottom of the DMR (around and below the ring header). As a result, a decision was made to design and test a new fluidized gas distributor. Conceptual designs focused on three different distributors: a modified design of the current system; a flat plate with bubble caps; and a conical bottom with dual plena. After testing the conceptual prototypes at Hazen Research Inc., the conical bottom distributor was chosen. A 30-day test with simulant SBW feed was then performed with the new distributor. During testing, a temperature differential of less than 2 deg. C was maintained in the DMR fluidized bed. The results supported the conclusion that poor fluidized gas distribution was the primary cause of high temperature differentials in the DMR fluidized bed. (authors)

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

Advancement of Commercial Intumescent Expanding Foams for Deactivation and Decommissioning in the Nuclear Sector - 20198

Florida International University (FIU), in collaboration with The Department of Energy's Office of Environmental Management (DOE-EM), Savannah River National Laboratory (SRNL), and sites across the Savannah River complex, have identified an operational requirement for a fixative technology that is intended to immobilize and/or isolate residual contamination within a 3-dimensional space. Fixation of radiological contamination can reduce worker risk and mitigate potentially hazardous conditions, however nearly every marketed contamination fixative has been found to be flammable; a significant concern in radiological facilities. Coupled with this, industry fixatives are normally used as a thin coating which can present problems when attempting to stabilize irregular geometry or areas that are difficult to access whilst ensuring full coverage. The technical evaluation and advancement of commercial-off-the-shelf (COTS) polyurethane foams has yielded a down-selected candidate that shows potential in meeting the requirements to support deactivation and decommissioning activities. Several performance criteria have been established and tested to progress the technology readiness level towards an active field demonstration (TRL-7). Such criteria include: mechanical failure limits, adhesive and cohesive properties, thermal/fire resilience, determining thermal behavior, ability to immobilize contamination, and a means of non-destructive evaluation of applications. The test scenario examined was targeted towards an application for decommissioning nuclear pipework, in which the down-selected polyurethane foam would act as a barrier to segregate pipework and mitigate the potential for release during cutting, packaging, and storage operations. Testing carried out at SRNL included: mechanical evaluation of tensile, compressive, and adhesion strength by dynamic mechanical analysis (DMA), as well as thermogravimetric analysis (TGA). FIU examined the foam's fixative properties by utilizing phosphorescent europium-dysprosium doped strontium aluminate powder to investigating the extent to which contamination can be immobilized. FIU has also exploited previous successes in the field of intumescent technologies to assess the down-selected foam's tolerance to an extreme fire scenario, while maintaining the ability to effectively mitigate a contamination release. Parallel to this, extensive thermal investigations were carried out to determine the upper boundary of anticipated heat generation during the curing process as heat generation has the potential to compromise rubber parts of contaminated enclosures. These investigations subsequently yielded a promising method for a non-destructive application evaluation by means of infrared thermography. Utilizing the high sensitivity of modern IR cameras, coupled with the heat generated during the curing process of the polyurethane foam, FIU has been exploring the concept of monitoring the external pipe surface for indications of an irregular or abnormal application, thus informing operational decision making. The testing carried out utilized several current 'best fit' ASTM standards, which serve as helpful guidelines for testing, however, a precise definition of the operational parameters and requirements is still necessary. With continued collaboration with SRNL, FIU aims refine said definitions and develop new standards by which this, and other decommissioning technologies, can be accredited by relevant standards based testing. (authors)

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

Evaluating Heat Extraction Performance of Closed-Loop Geothermal Systems with Thermally Conductive Enhancements in Conduction-Only Reservoirs: Preprint

We investigated the impact of disc-shaped and linear thermal enhancements on performance of closed-loop geothermal systems using steady-state and transient COMSOL numerical simulations. Thermal enhancements refer to thermally conductive material introduced in the rocks surrounding a wellbore to compensate for the relatively low rock thermal conductivity and increase heat extraction. Materials proposed include composite cements with thermally conductive additives such as metals or graphite to obtain overall cement thermal conductivities of 10 to 100x the rock thermal conductivity. Our simulation approach was modeling a subsegment (e.g., 100 m long) of a wellbore with idealized (e.g., perfect disc-shaped) thermal enhancements, and we report results as relative increase in heat extraction with respect to systems without thermal enhancements. We did not evaluate technical feasibility or cost of installing thermal enhancements. Simulations indicate that for 5-mm thick disc-shaped thermal enhancements with 5-m radius, repeated every 1 m along the wellbore and with thermal conductivity of 100x the rock thermal conductivity, the thermal output increases roughly 20% with respect to a closed-loop system without thermal enhancements. For fishbone structure thermal enhancements, we estimate a roughly 7% increase in thermal output for 5-m long radially outward pipes with half the radius of the main wellbore, repeated every 2.5 m, with thermal conductivity of 100x the rock thermal conductivity.

advanced geothermal systems↗