SEARCH · Engineering Papers

Results for “Nuclear Facility Components”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

SourceResearch connectionFrom yearThrough year

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 199 records · Page 11

All sources NASA NTRS DOE OSTI

DOE OSTI2020

Development and Test of High Temperature Surface Acoustic Wave Gas Sensors

The demand for sensors in hostile environments, such as power plant environments, aerospace environments, oil and gas extraction, and high-temperature metallurgy environments, has risen over the past decades in a continuous attempt to increase process control, improve energy and process efficiency in production, reduce operational and maintenance costs, increase safety, and perform condition-based maintenance in equipment and structures operating in high-temperature harsh-environment conditions. The increased reliability, improved performance, and development of new sensors and networks with a multitude of components, especially wireless networks, are the target for operation in harsh environments. Gas sensors, in particular H2 sensors, operating above 200 C are required in the instrumentation, process control and general safety of a number of industries including coal, natural gas, and nuclear power generation facilities, the aerospace and automotive industries, metallurgical production and defense-related applications. The surface acoustic wave (SAW) platform is a particularly promising option for high-temperature, harsh-environment gas sensing applications since the platform exhibits advantages, such as battery-free and wireless operation, small size, possibility for scale production using well-developed technologies from the semiconductor industry, and low cost of installation and operation. In this work, one-port SAW resonators (SAWRs) operating along five different orientations on a commercially available langasite (LGS) wafer employing Pt-Al2O3 electrodes and reflectors were designed, fabricated, and used as high-temperature H2 sensors. Two of the selected orientations were predicted and confirmed to have temperature-compensated operation above 150 C. A gas sensor test setup was developed, capable of gas cycling between N2, O2 and N2/H2 mixtures under extended high-temperature periods (up to 650 C for over 20 hours). Thin film Pt-Al2O3 was used as the electrode material for the transducers and reflectors capable of high-temperature operation, and also as H2 sensing film. In addition, yttria-stabilized zirconia (YSZ) thin films with Pt decoration were tested as sensing films aimed to enhance the SAWR sensor response to H2. The SAW devices were monitored in excess of 1700 hours in real-time during gas cycling sequences up to 600 C, leading to the following findings: i) the Pt-Al2O3 electrodes performed better for H2 sensing than the Pt-decorated YSZ sensing film, showing as much as 50% higher frequency variation response in the 200 C to 400 C range; ii) different crystallographic orientations operating on the same LGS wafer experienced different responses to H2 exposures up to 500 C; iii) the surface oxidation state of the SAWR sensors was shown to have an important impact on subsequent H2 exposure responses. In addition, a sensor system employing two LGS SAWRs, aligned along two different orientations, has been developed to simultaneously determine H2 presence and temperature. Finally, wireless interrogation of a SAWR sensor was successful within the gas cycling test fixture, and successful wireless H2 detection was achieved above 400 C.

Ayes Moncada, Armando↗

DOE OSTI2021

Cost- and Risk-Based Seismic Design Optimization of Nuclear Power Plant Safety Systems

Seismic analysis, design, and qualification of systems, structures, and components (SSCs) is a significant contributor to the capital cost of a nuclear power plant. To reduce capital costs of advanced nuclear power plants and make commercial nuclear energy more competitive, innovations are needed in their structural design and construction, and not just in the reactor core and associated systems. Seismic isolation has been identified as an important cost-cutting technology that enables standardization of equipment across various sites. This paper develops and demonstrates a cost- and risk-based seismic design optimization of a representative safety system in a nuclear power plant with the dual goals of minimizing overnight capital cost and meeting safety goals. The design optimization can also include component seismic isolation, in which case, the optimized design includes a set of equipment that needs to be seismically isolated to minimize capital cost. The open-source codes MASTODON and Dakota are used for seismic probabilistic risk assessment and design optimization, respectively. A generic nuclear facility with a safety system comprising SSCs that are common to nuclear power plants is considered for the demonstration of the design optimization and is assumed to be located at the Idaho National Laboratory site. Generic costs and seismic design cost functions are assumed for the SSCs of the safety system. The sum of the costs of the SSCs is minimized in the optimization process, while the risk of failure of the safety system is provided as a constraint. Furthermore, results show that the optimization process reduces capital costs significantly while automatically prioritizing the safety of SSCs that contribute most to the risk of the safety system.

42 ENGINEERING↗

DOE OSTI2024

Integrated DM1200 Melter Testing of HLW AZ-102 Composition Using Bubblers (Final Report)

This report documents melter and off-gas performance results obtained on the DM1200 HLW Pilot Melter during processing of simulated HLW AZ-102 feed. The principal objectives of the DM1200 melter testing were to determine the achievable glass production rates for simulated HLW AZ-102 feed; determine the effect of bubbling rate on production rate; characterize melter off-gas emissions; characterize the performance of the prototypical off-gas system components as well as their integrated performance; characterize the feed, glass product, and off-gas effluents; and to perform pre- and post test inspections of system components.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2024

DM1200 Tests with C-104/AY-101 HLW Simulants (Final Report)

This report documents melter and off-gas performance results obtained on the DM1200 HLW Pilot Melter during processing of simulated HLW C-104/AY-101 feed. The principal objectives of the DM1200 melter testing were to determine the achievable glass production rates for simulated HLW C-104/AY-101 feed; determine the effect of bubbling rate on production rate; characterize melter off-gas emissions; characterize the performance of the prototypical off-gas system components as well as their integrated performance; characterize the feed, glass product, and off-gas effluents; and to perform pre- and post test inspections of system components.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

NASA NTRS2014

Nuclear Cryogenic Propulsion Stage Affordable Development Strategy

The development of nuclear power for space use in nuclear thermal propulsion (NTP) systems will involve significant expenditures of funds and require major technology development efforts. The development effort must be economically viable yet sufficient to validate the systems designed. Efforts are underway within the National Aeronautics and Space Administration's (NASA) Nuclear Cryogenic Propulsion Stage Project (NCPS) to study what a viable program would entail. The study will produce an integrated schedule, cost estimate and technology development plan. This will include the evaluation of various options for test facilities, types of testing and use of the engine, components, and technology developed. A "Human Rating" approach will also be developed and factored into the schedule, budget and technology development approach.

Doughty, Glen E.↗

DOE OSTI2020

Development of Remote Cutting Tools for use at ATR - 20438

As a result of years of operations of the Advanced Test Reactor (ATR) at the Idaho National Lab (INL), the ATR canal has become congested due to the storage of materials including waste consisting of highly irradiated remote handled (RH) metals. In order to support its continuing nuclear mission the ATR requires canal space to support fuel storage and future core internal change out (CIC) components. INL staff determined that the canal must be cleaned out and waste removed to efficiently utilize the valuable canal floor space. It was determined that a methodology for processing, packaging, characterizing, and removing current and future irradiated hardware from the ATR canal is needed. Orano Federal Services (OFS) and its subcontractor Babcock Services Inc. (BSI) were selected to provide remote tooling in order to safely and efficiently size reduce the material currently stored in the ATR canal. INL developed detailed requirements that established the design and performance requirements for the tooling. Orano TN and BSI had recently successfully completed similar tasks in clearing highly activated materials from spent fuel pools and canals at commercial nuclear plants. That experience directly translated into a proven approach that could be used at the ATR thereby saving time and money and lowering risks. Based on BSI's demonstrated proficiency in the design and deployment of remote size reduction specialty tooling at comparable commercial facilities, Orano Federal Services was confident that a solution to INL's requirements could be delivered. The tooling had to be designed to be compatible with the operational requirements of the ATR canal as well as the functionality needed by the facility operators that will use the equipment. Some of the major design requirements and functions of the equipment included remote operations, capability to shear highly irradiated sections of aluminum and steel tubing, pipe, and components ranging in size from 0.6 cm to 16.5 cm diameter into 1.5 m -1.8 m lengths, utilizing buoyancy compensation to assist with underwater operations, and use commercially available existing technology and components as much as practical. The design team evaluated and selected commercially available equipment that could be modified to meet the design and performance specifications. In addition, a waste sizing table (WST) was designed and fabricated to further ensure the size reduction equipment would perform as expected and give the facility operators a stable and usable remote, underwater work platform. After successful development of the tooling, a mock up test and operator training was conducted to ensure the tooling performed as expected and met the requirements of INL. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2024

MARVEL Reactor Fuel Performance Report (Rev.2)

The Microreactor Applications Research Validation and EvaLuation (MARVEL) project is producing a high temperature liquid metal-cooled nuclear test bed at Idaho National Laboratory (INL) to ultimately improve the integration of microreactors to end-user applications. This ambitious effort seeks to design, authorize, construct, test, and operate the reactor within five years. In order to construct and operate the MARVEL reactor in a timely manner, the system will utilize materials and component designs which have already been used, qualified, or licensed from previous reactors. The MARVEL reactor will be located at the INL Transient Reactor Test (TREAT) facility in the north high-bay equipment pit and will use the existing 304 stainless steel-clad U-ZrH1.6 pin-type fuel system developed by General Atomics and purchased from TRIGA International. This fuel has been previously qualified under the United States Department of Energy’s (US DOE) Reduced Enrichment for Research and Test Reactors (RERTR) Program. Even though the regulator of the MARVEL reactor is the US DOE, the standards and overall approach recommended by the Nuclear Regulatory Commission is well-defined and utilized here. Following NUREG-1537 regulatory guidance, this report documents the authorization case for the MARVEL fuel system’s application to MARVEL and establishes stable and predictable fuel performance during the most thermophysically unfavorable conditions achievable in the MARVEL reactor. To that end, this report provides a comprehensive survey of the known thermophysical properties, performance, and quantitative relationships associated with the MARVEL reactor fuel element and uses this information to determine its mechanical integrity and risk of reaching unacceptable conditions during the most extreme accident scenarios predicted for the reactor using the most conservative assumptions available. The information contained herein is compiled from a combination of historical reports and peer reviewed scientific publication manuscripts. Known mechanisms under which the fuel is susceptible to failure are highlighted and compared to conditions that could exist in the MARVEL reactor during an unanticipated transient or accident scenario. The two scenarios considered for analysis in this report are (1) an unprotected loss of flow accident and (2) a hypothetical unprotected loss of coolant accident during the loss of flow accident. Preliminary 2D steady-state analyses herein indicate that both fuel-cladding chemical interactions and fuel-cladding mechanical interactions are negligible throughout the fuel’s operational cycle under both normal and high temperature accident scenario conditions. Although higher fidelity 3D time-dependent modeling and simulations are planned, the following may be concluded presently. The MARVEL fuel element maintains its geometric stability and structural integrity during the most extreme accident scenarios predicted for the MARVEL reactor. The hoop stress during the unprotected loss of flow accident reaches about -1.27 MPa; this negative stress indicates that it is compressive rather than tensile. The compressive stress is a result of the NaK pressure on the outside of the fuel element, caused by the restricted thermal expansion of the NaK coolant, exceeding the internal pressures generated inside of the fuel element. The hoop stress generated in the cladding during the unprotected loss of coolant accident reaches a maximum of about approximately 10 MPa, which is nearly an order of magnitude less than the predicted yield strength of the cladding under high-temperature accident scenario conditions. Calculations were compared with results from high performance computational simulations using BISON and are in very close agreement. A conservative MARVEL fuel meat peak temperature limit of 900 °C is recommended presently, which is about 180 °C higher than the peak fuel temperature predicted to occur during the most extreme accident. Based on the known properties and behavior of the MARVEL fuel element, the fuel successfully meets its design and safety requirements under normal and most extreme accident conditions with a large safety margin.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2026

MARVEL Reactor Fuel Performance Report

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

NRIC Annual Report FY 2025

The National Reactor Innovation Center (NRIC), established in August 2019, is a national United States (U.S.) Department of Energy (DOE) program. NRIC’s mission is to partner with industry and national laboratories to bridge the gap between concept, demonstration, and commercialization of advanced nuclear technology. NRIC accomplishes this through building or enhancing existing DOE infrastructure to support testing of components and systems that are key to successfully deploying advanced nuclear technology. NRIC’s vision is that by 2028, NRIC will be partnered with industry and accelerating the demonstration and deployment of advanced nuclear technology using DOE national laboratory infrastructure and expertise. NRIC will establish four new experimental facilities and two large reactor test beds for integrated technology demonstrations and experimentation by 2028 and complete two advanced nuclear technology tests by 2030. Achieving this vision will enable urgently needed abundant and affordable clean energy both domestically and internationally. NRIC’s success will inspire our nation and the global community to embrace the promising contribution of innovative nuclear reactor technologies to the clean energy economy and re-establish the U.S. as the global leader in advanced nuclear energy. NRIC is tasked with expediting the development of advanced nuclear energy technologies by bringing together private-sector technology developers and the world-class capabilities of the DOE national laboratory system. Through this program, the U.S. private sector is given access to the physical infrastructure available at DOE national laboratories to test and demonstrate their reactor concepts. NRIC works closely with the DOE-Nuclear Energy (NE) program that grants access to technical, regulatory, and financial support for commercializing nuclear energy. NRIC builds upon these new reactor concepts and technology successes to effectively strengthen U.S. nuclear leadership.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗

NASA NTRS1995

Preliminary results of radiation measurements on EURECA

The eleven-month duration of the EURECA mission allows long term radiation effects to be studied similarly to those of the Long Duration Exposure Facility (LDEF). Basic data can be generated for projections of crew doses and electronic and computer reliability on spacecraft missions. A radiation experiment has been designed for EURECA which uses passive integrating detectors to measure average radiation levels. The components include a Trackoscope, which employs fourteen plastic nuclear track detector (PNTD) stacks to measure the angular dependence of LET (greater than or equal to 6 keV/microns) radiation. Also included are TLD's for total absorbed doses, thermal/resonance neutron detectors (TRND's) for low energy neutron fluences and a thick PNTD stack for depth dependence measurements. LET spectra are derived from the PNTD measurements. Preliminary TLD results from seven levels within the detector array show that integrated doses inside the flight canister varied from 18.8 plus or minus 0.6 cGy to 38.9 plus or minus 1.2 cGy. The TLD's oriented toward the least shielded direction averaged 53 percent higher in dose than those oriented away from the least shielded direction (minimum shielding toward the least shielded direction varied from 1.13 to 7.9 g/cm(exp 2), Al equivalent). The maximum dose rate on EURECA (1.16 mGy/day) was 37 percent of the maximum measured on LDEF and dose rates at all depths were less than measured on LDEF. The shielding external to the flight canister covered a greater solid angle about the canister than in the LDEF experiments.

Benton, E. V.↗

NASA NTRS1995

Preliminary results of radiation measurements on EURECA

The eleven-month duration of the EURECA mission allows long-term radiation effects to be studied similarly to those of the Long Duration Exposure Facility (LDEF). Basic data can be generated for projections to crew doses and electronic and computer reliability on spacecraft missions. A radiation experiment has been designed for EURECA which uses passive integrating detectors to measure average radiation levels. The components include a Trackoscope, which employs fourteen plastic nuclear track detector (PNTD) stacks to measure the angular dependence of high LET (greater than or equal to 6 keV/micro m) radiation. Also included are TLD's for total absorbed doses, thermal/resonance neutron detectors (TRND's) for low energy neutron fluences and a thick PNTD stack for depth dependence measurements. LET spectra are derived from the PNTD measurements. Preliminary TLD results from seven levels within the detector array show that integrated does inside the flight canister varied from 18.8 +/- 0.6 cGy to 38.9 +/- 1.2 cGy. The TLD's oriented toward the least shielded direction averaged 53% higher in dose than those oriented away from the least shielded direction (minimum shielding toward the least shielded direction varied from 1.13 to 7.9 g/cm(exp 2), Al equivalent). The maximum dose rate on EURECA (1.16 mGy/day) was 37% of the maximum measured on LDEF and dose rates at all depths were less than measured on LDEF. The shielding external to the flight canister covered a greater solid angle about the canister than the LDEF experiments.

Benton, E. V.↗

DOE OSTI2024

Planning for Material Control and Accountancy at Liquid Fueled Molten Salt Reactors

The purpose of this report is to provide molten salt reactor (MSR) developers and future US Nuclear Regulatory Commission (NRC) license applicants with recommendations for developing an effective and practical material control and accounting (MC&A) plan, focused primarily on MSR designs that use circulating liquid fuel. Because of the breadth of MSR designs, there is no single, generic, detailed MC&A plan that will work for every design. The wide variation of fresh fuel salts, the method and frequency of loading fresh fuel, the reactor system design components (e.g., tanks, filtration systems, chemical processing streams), and waste streams will determine the specific measurement locations and instrumentation that can best meet MC&A objectives throughout an MSR facility. Additionally, MSR designs are rapidly evolving, and new design features and deployment scenarios that will affect MC&A are being explored and pursued. This report defines a generic MC&A approach that was developed for terrestrial (as opposed to maritime) deployments to meet the intent of NRC domestic safeguards and MC&A. MSR license applicants should consider nuclear safeguards (both domestic and international) and security throughout the design, as early as the preconceptual design phase. MC&A of special nuclear material (SNM) is an aspect of the NRC’s domestic safeguards program, alongside physical protection. Because liquid-fueled MSRs are reactors with SNM in nondiscrete (or item) form, it is likely that the NRC may require liquid-fueled MSR license applicants to submit a formal MC&A plan as a part of their license application. Currently, the NRC licensing protocol presents a challenge because the NRC MC&A regulations have not been updated to accommodate advanced reactors, including types of MSRs. Because no liquid-fueled MSR has been licensed for operation at the time of this report, no template or precedence for a successfully licensed MSR MC&A plan exists. However, the MSR license applicant can take advantage of the NRC’s published commitments to performance-based regulations. The authors recommend that the license applicant, or MSR designers, develop an MC&A plan throughout the design lifecycle and plan to submit a detailed MC&A program description, or MC&A plan, to the NRC as a part of a license application. No MC&A plan template or guidance exists that is specific to liquid-fueled MSRs. The authors recommend that license applicants discuss the topic of MC&A during preapplication engagement. Because of the uniqueness of MC&A for liquid fueled MSRs, the authors recommend that liquid fueled MSR developers engage with the NRC on the topic of MC&A in the early phases of its design development and follow up any time there are significant modifications in design plans that would affect MC&A. For example, topics like modifications in fuel handling processes, changes in uranium enrichment, or additional chemical processing streams added to the design could be discussed with the NRC specifically on the topic of MC&A.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Mild Acetylation and Solubilization of Ground Whole Plant Cell Walls in EmimAc: A Method for Solution-State NMR in DMSO-d6

Lignocellulosic biomass is mainly composed of polysaccharides and lignin. The complexity and diversity of the plant cell wall polymers makes it difficult to isolate the components in pure form for characterization. Many current approaches to analyzing lignocellulose structure, which involve sequential extraction and characterization of the resulting fractions, are time-consuming and labor-intensive. The present study describes a new and facile system for rationally derivatizing and dissolving coarsely ground plant cell wall materials. Using ionic liquids (EmimAc) and dichloroacetyl chloride as a solvent/reagent produced mildly acetylated whole cell walls without significant degradation. The acetylated products were soluble in DMSO-d6 from which they can be characterized by solution-state two-dimensional nuclear magnetic resonance (2D NMR) spectrometry. A distinct advantage of the procedure is that it realizes the dissolution of whole lignocellulosic materials without requiring harsh ball-milling, thereby allowing the acquisition of high-resolution 2D NMR spectra to revealing structural details of the main components (lignin and polysaccharides). The method is therefore beneficial to understanding the composition and structure of biomass aimed at its improved utilization.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2026

A machine learning based approach to online electron reconstruction at CLAS12

Online reconstruction is key for monitoring purposes and real time analysis in High Energy and Nuclear Physics experiments. A necessary component of reconstruction algorithms is particle identification that combines information left by a particle passing through several detector components to identify the particle’s type. Of particular interest to electro-production Nuclear Physics experiments such as CLAS12 is electron identification which is used to trigger data recording. A machine learning approach was developed for CLAS12 to reconstruct and identify electrons by combining raw signals at the data acquisition level from several detector components. Here, this approach achieves an electron identification purity above 75% whilst retaining an efficiency close to 100%. The machine learning tools are capable of running at high rates exceeding the data acquisition rates and will allow electron reconstruction in real-time. This work enhances online analyses and monitoring and can contribute to improved triggering at CLAS12. This machine learning driven approach will also be crucial for experiments aiming to transition to streaming readout operations where online reconstruction will be a key component of the data taking paradigm.

Artificial intelligence↗

DOE OSTI2023

Final Design for Thermal/Epithermal eXperiments (TEX) with Chloride Absorbers to Provide Validation Benchmarks for Y-12 Electrorefining Facility

Uranium electrorefining operations at Y-12 require validation for chlorine absorption and reflection. Plutonium chloride solution operations at Los Alamos National Laboratory require chlorine absorption validation. The growing need for chlorine validation is evident to the wider criticality safety community, with multiple attendees at the recent TEX 2.0 meeting at Lawrence Livermore National Laboratory (May 2023) requesting validation for chlorine (Idaho National Laboratory (INL), Institute de radioprotection et de surete nucleaire (IRSN), Savannah River Nuclear Site (SRNS), LANL, and Y-12). Los Alamos National Laboratory has recently performed, and benchmarked, an experiment titled Chlorine Worth Study (CWS) for the internal operations at the lab, but due to the difficulties in precisely characterizing the material compositions of the chlorine absorbers it is advantageous to perform a complimentary study with a different chlorine-based absorber material. Furthermore, having a uranium-based vs plutonium-based experiment provides a separate and important validation basis for criticality safety and nuclear data evaluation. The original final design report for the study of chlorine absorption using the TEX-HEU experimental base was presented in 2022, but used the same chlorine-bearing materials that were found to be difficult to characterize in the LANL benchmark. A complete redesign of the experiment has been performed looking at alternative absorber materials in various forms to produce an experiment that is fully characterizable. This report presents five novel chlorine experiments using the TEX-HEU test bed which provides direct comparison to the Y-12 and INL/Terrapower application needs, utilizing sodium chloride (NaCl) absorber plates. The absorber plates will consist of granulated NaCl (≥99.5% pure), which will be fully encapsulated in aluminum tins, providing a simple but effective chlorine-based absorber material that can be completely characterized. Of the five configurations presented in this report, it is expected that two or three configurations will be down selected for the actual experiment, with the other configurations being alternates. Three of the configurations are in the standard configuration, where the absorber is placed directly on the HEU fuel plates, and two in the sandwich configuration, where the absorber is surrounded by polyethylene moderators to force additional neutron thermalization prior to reaching the absorber. There are two thicknesses of NaCl absorber plates: 3/16” and 1/4” active thicknesses (i.e. not including the encapsulation). Both variations of the absorbers have an active absorber radius of 6” and a total radius of 7.5” to match the diameter of the HEU plates, with the outer 1.5” being aluminum encapsulation. The high-density polyethylene (HDPE) moderators are of the thicknesses: 27/16”, 7/4”, 1/8”, 11/16”, and 3/4”. The final configurations have six (one sandwich and one standard configuration), eight (one sandwich and one standard configuration), and 18 total fuel layers (standard configuration). The standard and sandwich configurations were designed such that the differences in moderator thicknesses are 1” HDPE, which were already procured for the original CED-2. The proposed configurations were precisely tuned to closely match the sensitivity profiles and neutron spectra of the Y-12 upset cases and were also compared to the INL/Terrapower upset cases. The assessment of experimental uncertainties of the non-absorber components was predicted to be 0.00114 Δk eff . The assessment of uncertainties resulting from the absorbers was predicted to be 0.00029 Δk eff . This results in a total uncertainty of 0.00118 Δk eff . Many of the largest uncertainties, namely the moderator densities, may be reduced with precision dimensional inspection of the components. The 1” HDPE moderators as well as the HDPE reflectors from the original CED-2 were incorporated in the final designs presented here. Additional HDPE moderator plates must be fabricated to complete the configurations. NaCl absorber plates will by fabricated at LLNL. The total additional cost is expected to be $\$$54,250 for the remaining components. Precise inspection, including dimensional, mass, density, and impurity, is recommended for all components. LLNL estimates that these costs are around $\$$12,000. It is expected, based on previous TEX-HEU experiments, that three weeks of experimental facility time is needed to complete the experiments.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2023

Code Benchmark of the HTTF Pressurized Conduction Cooldown Test Using SAM

The High Temperature Test Facility (HTTF) at Oregon State University is an integral system test facility to simulate postulated reactor transients of prismatic high-temperature gas-cooled reactors(HTGRs). A series of test campaigns was launched, providing abundant test data that could be used to benchmark reactor system analysis codes like the System Analysis Module (SAM). In this study, a SAM model of the facility is developed based on the two-dimensional (2D) ring model approach. All components including the ceramic matrix, graphite heaters, helium coolant channels, core barrel, upcomer, pressure vessel, and reactor cavity cooling system are modeled as concentric cylindrical rings. The model is used to simulate one of the benchmark problems-Pressurized Conduction Cooldown (PCC)-within the scope of the Organisation for Economic Co-operation and Development Nuclear Energy Agency International HTTF Benchmark. The simulations consist of two parts. In the first part, operating and boundary conditions as well as thermophysical properties of materials are specified for the benchmark problem. In this work, results from the first part will be used in code-to-code comparison. In the second part, the SAM model is used to simulate Test PG-27, which is the first PCC test carried out in the HTTF, with only two of the ten heater banks activated. The results in the second part are used for code-to-data comparison. Because the helium coolant flow rate is not measured in this facility, it is estimated using the input power and inlet/outlet coolant temperatures. Additionally, radial heat flow in the ceramic blocks is complicated by hundreds of cylindrical coolant channels and heater rods embedded in them. As such, it is necessary to deduce an effective thermal conductivity for the ceramic to analyze the core thermal behavior. SAM predictions of the helium coolant and ceramic temperatures are compared with test data measured in three equivalent sectors. Overall, the SAM results agree reasonably well with test data within the variation of data among the three sectors, which demonstrates SAM's capability in capturing transient effects in HTGR using the simplified 2D ring model.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2020

Treatment of Problematic Reactive Metal Wastes Using the GeoMelt{sup R} In-Container Vitrification (ICV{sup TM}) Process - 20326

Decommissioning of sodium-cooled reactors and fast reactor technologies has generated a number of reactive metal waste configurations that are problematic to treat and typically lack cost effective treatment methods and disposition options. As a result, Veolia Nuclear Solutions, under contract with Idaho National Laboratory (owned by the U.S. Department of Energy and managed and operated by Battelle Energy Alliance, LLC) demonstrated its GeoMelt{sup R} In-Container Vitrification (ICV){sup TM} technology to safely convert sodium metal to a non-reactive vitrified oxide form. The demonstration project, supported by glass formulation and crucible testing, consisted of a series of ICV{sup TM} melts that processed elemental sodium into stable non-reactive glass. INL is currently implementing GeoMelt{sup R} technology as a means to safely and reliably convert radioactive reactive metal residues that contaminate sodium cooled reactor components into waste forms that comply with existing disposition pathways. Reactive metal wastes require treatment in order to remove the Resource Conservation and Recovery Act (RCRA) reactivity and ignitability characteristics to comply with land disposal restrictions. GeoMelt{sup R}, which is an alternative to other potential treatment approaches, provides a robust approach that chemically converts the reactive metals to an inert oxide while also immobilizing radionuclides in a vitrified waste form with durability equal to or better than vitrified nuclear fuel reprocessing wastes (very robust and inert waste forms). Most other treatment approaches generate hydrogen gas which is problematic. In 2016, Veolia Nuclear Solutions first demonstrated the effectiveness of the GeoMelt{sup R} ICV{sup TM} process in deactivating reactive sodium metal. Crucible, bench-scale, and engineering-scale demonstrations were conducted on several surrogate waste configurations with various ratios of sodium metal and glass formers. Each ratio and configuration demonstrated complete deactivation of the surrogate sodium metal. Follow-on work in 2017 demonstrated the deactivation of reactive sodium by GeoMelt{sup R} ICV{sup TM} at a higher waste loading relative to previously demonstrated work performed in 2016; the higher waste loading optimized glass chemistry while enhancing the economical full-scale treatment of reactive metals. Additionally, follow-on demonstration testing in 2018 and 2019 focused on more complex shapes and other reactive-metals (mocked up Experimental Breeder Reactor II [EBR-II] subassembly, sodium filled heat exchanger, and a can containing sodium potassium alloy) which were all performed at engineering scale. Veolia Nuclear Solutions designed, installed, and commissioned in September 2018, at Perma-Fix Northwest in Richland Washington, a 10-metric ton full-scale GeoMelt unit (GeoMelt{sup R} Richland) for the treatment of reactive metal wastes. As of September 2019, over 900 55-gallon drums containing a total of around 3,500 lb of sodium with low levels of radioactivity have been treated at GeoMelt{sup R} Richland, with resulting glass monoliths disposed at the Nevada National Security Site (NNSS). A full-scale radiological demonstration melt on an actual EBR-II subassembly has also been performed using the full-scale melter in 2019. The GeoMelt{sup R} technology is a proven radioactive waste treatment technology capable of immobilizing radioactive wastes, including bulk rubble such as drums and other steel vessels usually without pretreatment. Utilizing the GeoMelt{sup R} technology to treat reactive metals eliminates pretreatment steps resulting from having to separate the reactive metal from steel containers or jackets as GeoMelt{sup R} can easily operate at temperatures sufficient to melt the steel and expose the reactive metal for treatment. Eliminating handling steps of reactive metals is a significant safety advantage since reactive metals are pyrophoric. The results generated as a part of the 2018-2019 demonstration program are presented in the paper. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Methane generation rates from Tank 50 simulant with methylmercury containing saltstone grout solids

This report documents the tests performed to address the requirements listed in the Technical Assistance Request, which resulted in a Test Plan to obtain the required information. To that end the results are documented herein to determine the capacity, and reaction time, for methylmercury (MeHg) contained within Tank 50 salt solution simulant (T50SS) to produce methane when mixed with saltstone dry feed ingredients, as individual components (cement, slag cement, and fly ash) or a combination of those components.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗