Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “ASME”

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.

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

At least 793 records · Page 44
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Evaluating the Performance Envelope of Axial Flow HEPA Filter Designs - 17466

Nuclear grade high efficiency particulate air (HEPA) filters are routinely employed as the final control element in confinement ventilation systems. Approximately 6000 nuclear grade HEPA filters are purchased each year for use in the Department of Energy National Nuclear Security Administration (DOE/NNSA) complex. The traditional HEPA filter utilizes a pleated filter medium of fibrous glass that has been produced in a manner analogous to manufacturing paper. Containment ventilation systems are designed to condition the air-stream (temperature, relative humidity, moisture, particle size distribution of aerosols, and aerosol concentration) to levels and conditions that fibrous glass HEPA filters can handle. Axial Flow filters can be purchased with physical separators or embossed medium to maintain proper pleat geometry. Separator-less designs use media embossed with a corrugation pattern to retain pleat geometry. At least some versions of the separator-less filter packs have demonstrated unexpected and rapid failure under slightly elevated temperature and humidity conditions. The effects of age on HEPA filter performance have been a concern for at least two decades. A combination of physical evaluation of aged filter media and a review of filter failure records resulted in Appendix C of the Nuclear Air Cleaning Handbook (NACH), Determination of HEPA Filter Life. Mississippi State University is conducting studies to more clearly identify the operating envelope of separator and separator-less designs of ASME AG-1 Section FC axial flow HEPA filters under a set of generic design basis conditions. A parallel study is also being conducted to evaluate the effects of age on fibrous glass HEPA filters. Activities described in this paper have been conducted under Institute for Clean Energy Technology (ICET's) NQA-1 nuclear quality assurance program approved by DOE-EM. Testing meets the requirements of EM Quality Assurance Program (QAP) EM-QA-001, Rev. 1, June 11, 2012 that includes the applicable requirements of NQA-1-2008/2009a and DOE Order 414.1D. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2017

Software Quality Assurance for LeachXS/ORCHESTRA as a Tool for Assessment of Waste Management Systems Performance - 17486

Qualified modeling and simulation tools and experimental data are needed to provide a reasonable basis for the incorporation of realistic wasteform retention assumptions (i.e., those without orders-of-magnitude conservatism) into performance assessments and regulatory decisions. Use of highly conservative assumptions in performance assessments has historically led to the overprediction of release and near-field transport by as much as 4- to 6-orders of magnitude thereby limiting the nature and quantities of certain contaminants of concern in near surface disposal facilities. This project provides peer-reviewed models and characterization methods to provide more reasonable assumptions for use in performance assessments and regulatory decisions. Simulations tools have been developed to predict hydraulic and chemical performance of waste-forms and barriers used in nuclear applications under several important aging and degradation scenarios, including sulfate attack, carbonation, oxidation, cracking, and leaching; these tools are complaint with DOE Order 414.1D and ASME NQA-1. Wasteform characterization methods have been developed that are consistent with US EPA's Leaching Environmental Assessment Framework (LEAF) currently part of EPA's Hazardous Waste Test Methods (SW-846); the resulting experimental data are evaluated using LeachXS. ORCHESTRA models, that can be called seamlessly from LeachXS, have been developed representing both laboratory and field conditions; these models have undergone an extensive Software Quality Assurance process where they have been calibrated (where necessary), verified, and validated to be compliant with relevant guidelines. The models used for field predictions have also been incorporated into the Cementitious Barriers Partnership (CBP) Software Toolbox that allows probabilistic evaluation of field scenarios. The foci over the next year will be on providing 1) a mechanistic basis for projecting the retention of Tc-99, I-129, and other important radionuclides in near-surface disposal facilities at the Savannah River and Hanford Sites and 2) a technical basis for assessing the efficacy of macro-encapsulation of mercury-contaminated debris at the Oak Ridge Reservation. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Ensuring NQA-1 Compliance with Nuclear Next Gen Technologies Performing a Safety Function Associated with Autonomy or Remote Systems - 17549a

First-of-a-kind and cutting edge Next Gen technologies deployed in a nuclear environment may initially appear to fall outside the ability to be qualified in accordance with ASME NQA-1 for a nuclear safety function. However, NQA-1 compliance is possible especially if baked into the early stages of the technology development process. This is particularly applicable as part of the increasingly prevalent application of autonomous processes, advanced manipulation, man-machine interfaces, exoskeletons and other advanced and unique ability options. (authors)

42 ENGINEERING↗
DOE OSTI2017

Development of High Strength HEPA Filters for Use in Demanding Air and Gas Streams - 17644

High-efficiency particulate air (HEPA) filters are used in a wide variety of applications, including the removal of radioactive particulate from process off-gas streams and the venting of waste storage containers. Porvair Filtration Group, specializing in the design and manufacture of robust filters in numerous industries, has worked to develop HEPA filter designs for use in such high-demand nuclear process-gas and air streams. Using the requirements of section FK of the American Society of Mechanical Engineers AG-1 Code on Nuclear Air and Gas Treatment (ASME AG-1) as a guide, Porvair Filtration Group has designed filters which can withstand upset conditions to include elevated temperatures, high relative humidity, high particle-loading, and an unconventionally high differential pressure across the filter media; while maintaining excellent removal efficiencies. The process by which Porvair Filtration Group has reached optimized designs was based upon an extended test and development program. Design, manufacture, and testing of flat sheet media samples, geometrically equivalent model filters, and full-scale prototypes was performed at several Porvair Filtration Group facilities located in the United States and the United Kingdom. The methods and results of this development program are outlined in the following paper. (authors)

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

Success in commissioning a key U.S. radioactive waste burial site through the technical support of Dufrane Nuclear Shielding Incorporated (DNSI) - 15476

With the -phase out- of the South Carolina based Barnwell Radioactive Waste Burial Site, and the stockpile and storage of radioactive waste and irradiated components throughout the U.S. Nuclear Industry for decades, the opening of the Texas Low-Level Radioactive Waste Disposal Compact Facility by Waste Control Specialists (WCS), is a game changer in solving this industry-wide issue. For the first time in decades, generators now have the opportunity to send their waste to a site specifically designed to permanently sequester the waste in a licensed facility that will protect public health and the environment, better than any built to date. As history has shown through the many decades of the U.S. Commercial and Federal Nuclear Industry, site operators and management have relied heavily on contractors and vendors with exceptional expertise and decades of experience to achieve success and meet their goals. The commissioning of WCS was no exception as DNSI was contracted to design, develop, build and deploy a major component required for the opening and successful operation of the WCS site. This component, being the Modular Concrete Canister (MCC), is the -work horse- of the site. With its enormous size, weight and strength, it provides many functions that meet and comply with the site performance assessment (PA) under 10 CFR Part 61, the site WAC, and Texas Low-Level Radioactive Waste Disposal Authority. Offsite contracting the production and delivery of this critical component was a major step toward success in meeting schedule and cost for the much needed opening of the Texas Compact and Federal site. In selecting a shielding vendor, WCS partnered with a U.S. Nuclear Service company with over 30 years of experience which spans the entire U.S. Commercial, Federal, and Nuclear Industry. The task given DNSI was not a simple one, it involved every aspect from start to finish of the design, production and delivery of rectangular, cylindrical and -irradiated hardware- (IH) MCCs, that could accept waste up to 100 Sv /Hr contact dose rates. The enormity of this project was quite clear and required extensive quality control, construction of batch plants, purchase of huge quantities of raw materials, setting up a production site in a neighboring city, relocation of personnel from company headquarters, and safely transporting 8,000 Tons of MCC components 75 miles one-way on West Texas highways daily. Most important of all, this was accomplished safely with no incidents, no component failures or rejects, and on schedule. The start of receipt, of -in-compact waste- and -out-of-compact waste- was conducted virtually in a 3 month period time span. To help guarantee success and quality of product, all forms had surfaces machined, heavy duty elastomer coatings were applied to concrete surfaces, only galvanized anchors were used, and the batch plant was run to strict ASME standards thus testing every 50 cubic yards of concrete poured. With significant teamwork and success from Vendors such as DNSI, WCS and the Texas Compact has evolved as one of the leading radioactive waste sites in the U.S. and the world, providing safe, secure and compliant permanent waste disposal services by using technical expertise and innovation of the highest standards. (authors)

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

3-60B Type-B shipping cask enhanced operational features to facilitate leakage rate testing - 15585

EnergySolutions is fabricating a new 3-60B Type B Shipping Cask that will be ready for use in spring 2015, designed for shipping Type B quantities of activated and/or non-fuel-bearing reactor or accelerator components. EnergySolutions, in consult with Leak Testing Specialists, Inc., has implemented an improved design of the 3-60B cask closures to better facilitate periodic and maintenance leakage rate testing and to allow end users to more easily perform pre-shipment leakage rate testing to confirm proper assembly of the cask closures. A typical Type B cask closure lid design incorporates a double O-ring configuration, where the inner O-ring provides containment and the outer O-ring is used for leakage rate testing of the containment O-ring seal. Leakage rate testing of a double O-ring configuration is typically accomplished by filling the cask cavity with helium, pulling a vacuum on the annular space between the two O-rings and monitoring for helium leakage through the containment O-ring. One issue is the potential for helium saturation of the O-ring elastomeric material, which can result in leakage rate test failures. The cask cavity volume is typically large, and takes a significant amount of time to fill with helium. As a result, the saturation time for certain elastomeric compounds may be exceeded, leading to potential false-negative leakage rate testing results. To avoid this, that the cask cavity typically is loaded with sealed container(s) to reduce the free volume and the fill time. The improved 3-60B cask closure lid incorporates a triple O-ring seal configuration that facilitates maintenance leak rate testing of the closure lid containment O-ring seal and eliminates the issues discussed above. The triple O-ring configuration uses the middle O-ring as the containment seal, with the inner and outer test O-rings creating relatively small annular volumes for testing purposes. These small annular volumes are quick to fill, minimizing the issue of Helium saturation of the O-ring material. This seal configuration also makes it possible to detect and replace a damaged O-ring seal during loading operations, or dress a damaged sealing surface, and perform the required maintenance leakage rate testing of the replacement O-ring at the plant site without the need to unload the cask and send it to a maintenance facility. Also, a second test port has been added to each O-ring annulus to allow leakage rate testing to be accomplished using a formal system calibration that complies with the requirements and recommendations of ASME Section V, Article 10 [4], ANSI N14.5-97 [2], and ANSI N14.5-2014 [3]. This provides a simple, reliable, and accurate means to perform leakage rate testing with a test confidence level that cannot be achieved without the second test port. The improved closure and test port design simplifies and shortens the time required to perform leakage testing, resulting in reduced operating costs, greater schedule security, and dose savings for the cask users. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

The Publication and Endorsement of the ASME/ANS Probabilistic Risk Assessment Standard for Advanced Non-Light Water Reactor Nuclear Power Plants

Due to growing interest in non-light water reactor (NLWR) concepts, the predecessor to the American Nuclear Society (ANS)/American Society of Mechanical Engineers (ASME) Joint Committee on Nuclear Risk Management (JCNRM) initiated development of a NLWR probabilistic risk assessment (PRA) standard in 2006. The developed standard, released for trial use in 2013, was the first integral PRA standard created by the JCNRM, containing requirements ranging from the development of initiating events to the analysis of offsite radiological consequence. Based on feedback from pilot applications of the trial standard, a revised version of the standard was developed and balloted in 2020, with unanimous passage by the JCNRM and formal publication as an American National Standards Institute (ANSI) standard in 2021. The paper provides an overview of the standard and its development process. In addition, the paper reviews the latest status of U.S. Nuclear Regulatory Commission endorsement efforts through Regulatory Guide 1.247.

Fleming, K.↗
DOE OSTI2016

Corrosion of Candidate Materials in Molten FLiBe Salt for Fluoride-Salt-Cooled High-Temperature Reactor (FHR)

Molten fluoride salts have attracted considerable attention because of their desirable thermophysical properties that lead to superior heat transfer. Recently, fluoride salts are being considered as coolants in the fluoride salt-cooled high temperature reactors (FHRs). The successes of the molten salt reactor experiment (MSRE) program in 1950's-70's at Oak Ridge National Laboratory (ORNL) has also driven the pursuit of molten salt nuclear reactor as one of the lead next generation nuclear reactor concepts. Unlike MSRE where the uranium fuel was dissolved in the molten fluoride salt, the FHR concept is expected to use molten salt (2LiF-BeF{sub 2}, FLiBe) only as coolant in conjunction with solid TRISO fuel particles immersed in the molten salt. One of the important challenges in FHR development is corrosion of structural materials in high-temperature molten 2LiF-BeF{sub 2} (FLiBe) salt. Currently, 316 stainless steel (nominal composition: Fe-12Ni-17Cr-2.5Mo-1Si-2Mn-0.03C) and Hastelloy N (nominal composition: Ni-7Cr-16Mo-5Fe-1Si-0.8Mn) are deemed as promising candidate alloys for the construction of some of the structural components of FHR. 316 stainless steel is ASME code certified for high temperatures (up to 760 deg. C) applications. Hastelloy N has been shown to have outstanding corrosion resistance in molten FLiBe but present code certification data limit the use of this alloy to only about 704 deg. C. Modifications of Hastelloy N with higher creep strength are presently being developed at ORNL. Other materials such graphite, SiC/SiCf and C/Cf composites, are also considered for structural applications in the FHR. The presence of these multiple materials in a molten FLiBe salt medium can further exacerbate and complicate corrosion effects. Therefore, it is necessary to understand the corrosion behavior of these materials in molten FLiBe salt. The corrosion behavior of Hastelloy N, 316 stainless steel, and CVD SiC and two types of SiC-SiC{sub f} composites have been tested in molten FLiBe salt at 700 deg. C. When tested in metal-lined crucibles both Hastelloy N and 316 stainless steel showed very good corrosion resistance. Samples tested in graphite crucibles showed greater propensity for corrosion and carburization was observed in the near surface regions of the alloys. CVD SiC showed outstanding corrosion resistance. However, the two SiC-SiC{sub f} composites exhibited corrosion attack implying that the manufacturing process may play a role in the corrosion of these composites. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Development of a Model Correlating Axial HEPA Filter Performance Between Elevated and Standard Conditions

The ASME AG-1 Code specifies minimum rated airflow based on standard cubic feet per minute (SCFM) for axial high efficiency particulate air (HEPA) filters and actual cubic feet per minute (ACFM) for radial HEPA filters with corresponding maximum pressure drop. A quantitative analysis can be performed to demonstrate the variation in filter efficiency and develop a correlation of performance between axial filters subjected to ACFM and SCFM. This study illustrates the variation in filter efficiency of axial HEPA filters subjected to ACFM and SCFM flow rate conditions at varying atmospheric pressures. It also presents the methods used to calculate correction factors to convert filter efficiency and pressure drop from elevated to standard conditions. The Axial Large Scale Test Stand (ALSTS) in the Institute for Clean Energy Technology (ICET) at Mississippi State University has been utilized to control aerosol concentration, flowrate, atmospheric pressure, relative humidity, and temperature. The filter test matrix consists of axial HEPA filters rated at 1,000 and 1,500 CFM with varying filter pack types. Each filter is subjected to varying pressures of 14.70 psi, 13.55 psi, 12.40 psi, and 11.25 psi, temperatures of 70 deg. F, 110 deg. F, and 150 deg. F, and relative humidities of 50% and 80%, resulting in twenty-four unique testing points. The corresponding SCFM flow rates are computed for each point in order to form a comparison. Evaluation of the effects of designating ACFM or SCFM will be based upon the filter efficiencies of generated Dioctyl Phthalate (DOP) particles at a diameter of 0.3 micrometers. The results show filter efficiency decreasing as atmospheric pressure decreases due to increased filter face velocity. A correlation model is developed to convert filter efficiency and pressure drop at elevated conditions to those of standard pressure, relative humidity, and temperature.

42 ENGINEERING↗
DOE OSTI2020

Application of GOTHIC to Groundwater Transport Analysis - 20152

Migration of pollutants and hazardous wastes, potentially containing radioactive isotopes, via groundwater transport is a concern at most waste cleanup sites. Predictive analysis can be used to evaluate mitigating actions intended to minimize impact on the environment and public exposure. GOTHIC is a multipurpose thermal hydraulics code that is used extensively in the nuclear industry for design, licensing and operation evaluations. It combines the capabilities of typical one-dimensional system codes and the essential features of Computational Fluid Dynamics (CFD) codes for three-dimensional analysis. There are other codes that are specifically developed for groundwater transport analysis and the results presented here are consistent with prior analyses. However, GOTHIC has some unique features that offer advantages for applications related to nuclear waste. Most importantly, it has been developed and maintained under a Quality Assurance program in compliance with the requirements of 10CFR50 Appendix B [2] and applicable portions of ASME NQA-1 [3] since 1995. Available GOTHIC capabilities that make the code especially useful for groundwater transport of nuclear materials include: - Tracking of any number of tracer elements for contaminants and other species of interest; - Radioactive decay and progeny of tracer elements; - Adsorption/desorption of tracer elements; - Tracking of any number of dissolved gases; - Release and absorption for dissolved gases; - Vapor phase tracking; - Non-Newtonian fluid modeling. The general porous body modeling approach makes GOTHIC well suited to groundwater transport analysis. The multi-region modeling approach used by GOTHIC simplifies model construction for regions of varying hydrologic characteristics and focuses the computational effort on regions of particular interest while simultaneously capturing the macroscopic response and any feedback effects across the larger domain. The applicability of GOTHIC to groundwater transport applications is demonstrated by comparing code results with available analytic or semi-analytic solutions for groundwater behavior. (authors)

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

Tracking Radioactive Isotopes in HVAC and Application for Hot Cell Analyses - 20155

In recent years, the surge in the number of isotope production facilities under design has increased the need to analyze isotope migration in facility Heating, Ventilation, and Air Conditioning (HVAC) during normal operations and accident scenarios. These facilities are used to produce isotopes for medical, security, and industrial applications and as such are subject to license and regulatory requirements 10CFR20, 10CFR30, 10CFR50, and 10CFR70. GOTHIC, a general-purpose thermal-hydraulics software package, includes the ability to model isotopic tracers, radioactive decay and isotope migration as well as HEPA and charcoal filters for isotope retention. A GOTHIC model was developed by Zachry Nuclear Engineering (ZNE) to examine the effect of negative room pressure, HEPA filtration, and HVAC fluctuations on radiation areas, hot cells and gloveboxes. Radioactive tracers were used to simulate the concentration of spills within contaminated areas, track the migration of isotopes of interest, and determine the isotopic retention and buildup on facility HEPA filters. A variety of isotopes with concern to dose (e.g., Kr-85, Sr-90, I-131, etc.), including their decay and progeny, are included in the analysis. Negative pressures are maintained in the regions of interest by a representative central HVAC system equipped with a volumetric fan that exhausts to the environment after a series of isolation valves and HEPA filters. GOTHIC is an industry trusted tool for providing engineering solutions for a variety of applications, including fission product tracking, aerosol and particulate transport and ventilation assessments. The software provides an integrated analysis environment that includes a graphical user interface (GUI) for constructing analysis models, a numerical solver that includes parallel processing capabilities and a post-processor for evaluating simulation results. It solves the conservation equations for mass, momentum and energy for multicomponent, multi-phase flow in lumped parameter and multi-dimensional geometries (1, 2, or full 3D), including the effects of turbulence, diffusion and buoyancy. It has been developed and maintained under a Quality Assurance program in compliance with the requirements of 10CFR50 Appendix B and applicable portions of ASME NQA-1 since 1995. GOTHIC has been used for assessing both forced and natural convection conditions for a wide range of applications, including: - Tracking concentration of hazardous gases and chemicals for habitability and safety assessments - Determining ventilation and filtration requirements and optimizing location and arrangement of these systems - Room heat-up, including diverse and Flexible coping strategies for Extended Loss of AC Power (FLEX/ELAP) - Equipment Qualification (EQ) A distinctive feature of GOTHIC is the ability to track many different fields/substances in a simulation, including user defined tracer elements, in the liquid, vapor and droplet fields as well as surfaces and filters. This capability allows GOTHIC to model fission product transport and release or the removal of particulates or harmful toxins from exhaust gases using a spray scrubber or other types of filtration systems. GOTHIC also includes models for engineered equipment, such as fans, filters, charcoal filters, dryers/demisters, dampers, etc. The aerosols and other filtered material are removed or accumulated in these components. The range of aerosol and radiological applications that GOTHIC has been used for includes: - Source Term: Primary Coolant (Equilibrium) Activity; Non-Water Coolant Source/Leakage. - Conditions for Iodine Re-evolution: Sump/Suppression Pool Conditions and pH; - RWST Conditions and pH. - Isotope Removal Mechanisms: Containment Sprayed and Unsprayed Region Mixing; Charcoal Filter Heating due to Iodine decay. - Radionuclide Transport and Decay: Post-LOCA Release in containment; Transport between connected Compartments and vent systems; Groundwater transport of radionuclides. - Non-Newtonian Fluid modeling for sludge, waste tanks, etc. ADAMS ML071581053 (titled 'Best Practice Guidelines for the use of CFD in Nuclear Reactor Safety Applications') poses guidelines for applying single phase CFD codes in nuclear reactor safety problems and GOTHIC is listed as a 'tool for 3D flows' and 'dispersal and deposition of radionuclides.' The Nuclear Quality Assurance (NQA) pedigree of GOTHIC is an important aspect for applications in the nuclear industry. The fundamental tracer models (convective transport, molecular and turbulent diffusion, removal mechanisms, etc.) have been verified using analytical solutions and validated against applicable separate effects tests. Also, GOTHIC has been benchmarked to many integrated effects tests, including Phebus FP (Fission Product). GOTHIC gives good agreement for the buildup and decay of fission products in Phebus Test 3. The model developed by ZNE demonstrates GOTHIC's applicability and acceptability for use in analyzing the migration and retention of radioactive isotopes and their progeny in normal operation and accident scenario analyses for isotope production facilities, hot cells, and gloveboxes. The tracer activities calculated by GOTHIC can then be used in downstream radiation transport and shielding codes like RADTRAD-NAI{sup C}, MCNP{sup R}, and MicroShield{sup R} to determine on-site and of-site doses. (authors)

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

DOE Hanford Tank Side Cesium Removal System (TSCR) Update: Fabrication and Factory Acceptance Testing - 20300

The Tank Side Cesium Removal (TSCR) System is a modular, quickly-deployable, and passively safe system for efficiently removing solids and cesium from alkaline liquid waste raffinate created by fuel processing on the Hanford site. This paper provides an overview of the project with particular emphasis on completion of fabrication and factory acceptance testing (FAT). All TSCR System components are inside three enclosures known as the Process Enclosure, Ancillary Enclosure, and Control Enclosure. The Process Enclosure is seismic design category II (SDC-II) and built to withstand the rigors of transportation and lifting, high winds, volcanic ash fall, and heavy precipitation (snow) over a temperature range of -25 to 115 deg. F. The Process Enclosure and Control Enclosure were fabricated in AVANTech's Columbia, SC facility, while the Ancillary Enclosure was fabricated in AVANTech's Richland, WA facility. The design and preliminary documented safety analysis were completed in December 2019, with all work being performed in accordance with ASME NQA-1 quality assurance requirements. The Process Enclosure, Ion Exchange Columns (IXCs), and several other TSCR components have a nuclear safety function and are designated as safety significant structures, systems, and components (SS-SSCs). The NQA-1 Commercial Grade Dedication (CGD) process was used to identify critical characteristics in order to accept materials and items used for fabrication of several safety significant TSCR components. This paper also addresses unique aspects and challenges of implementing CGD in DOE facilities and systems. After fabrication and preliminary testing, the Process and Control Enclosures were shipped to AVANTech's Richland facility where the complete TSCR System was setup and connected in a represented layout and configuration that will be used on the Hanford site. Test tanks and pumps were put in place and connected to the TSCR Process Enclosure in a manner that mimicked interfacing Double Shell Tanks in the 241-AP Tank Farm. The TSCR System was then tested/operated under conditions equivalent to those planned for onsite operation. Testing was separated into five separate phases as defined below: - Phase 1: Confirmation of TSCR System readiness for the FAT; - Phase 2: Verification of TSCR System operation under standard conditions; - Phase 3: Demonstration of TSCR System operation under non-standard conditions; - Phase 4: Demonstration of ion exchange media removal from an IXC; and - Phase 5: Completion of TSCR System lift plans and site mobilization. The Hanford Tank Operations Contractor, Washington River Protection Solutions (WRPS), participated in all phases of testing. Testing at AVANTech's Richland Facility allowed WRPS personnel to familiarize themselves with TSCR operations in a radiologically clean facility prior to onsite mobilization. Engineering and operations personnel gained valuable experience that allowed them to better refine onsite operational protocols and procedures - including handling and movement of IXCs with a forklift. This paper describes the technologies used by the TSCR System as well as how these technologies are deployed and operated to advance the Hanford mission of safely, efficiently, and effectively treating tank waste. It also provides detailed information on fabrication and FAT, including operational data from simulated tank waste runs. A project schedule for WRPS' planned construction acceptance testing, operational acceptance testing, and actual tank waste treatment is provided. This presentation should be of great interest to parties responsible for the design, testing, and mobilization of modular NQA-1 technologies for the treatment of raffinate and similar liquid wastes containing high concentrations of radioactive cesium. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Improving plant efficiency through the installation of TecSOLCEP online monitoring system - 362

In February 2016, Nuclear Energy Institute (NEI) launched its 'Delivering the Nuclear Promise (DNP)' strategic plan with the objective of reducing electric generating costs by 30% across the nuclear power industry by 2018. Improving efficiency has been addressed as one of the strategic focus areas in the plan. In nuclear power plants (NPPs) there is room for improvement in main components efficiency and TecSOLCEP is a software tool that has proved its value for that purpose. TecSOLCEP is a software tool for monitoring, diagnosing and quantifying performance deviations in Power Plants. It faithfully follows the methodology of ASME standards. Over 15 years of success stories in more than 50 electric power plants of different technologies worldwide, justify the benefits for the plants. The implicit concept in TecSOLCEP is 'Benefit based Maintenance' (BbM) as complementary to usual maintenance strategies like time-based, condition-based or predictive. These maintenance strategies are oriented to obtain the Safety and Reliability goals shared by all the electricity generation technologies. BbM is oriented to get the maximum economic benefit by Efficiency care, which historically has not been a priority, especially in nuclear plants. BbM applied with the help of performance monitoring tools like TecSOLCEP, is called to play an important role to improve efficiency which, in plants mainly operating at full power, is equivalent to increase the maximum power. Nuclear power plants that intend to extend their lives may also take advantage of identifying and quantifying inefficiencies in equipment to better address necessary investments. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

University of Nevada, Reno Graduate Certificate in Nuclear Packaging

This paper describes a new Graduate Certificate in Nuclear Packaging (GCNP). Nuclear materials are currently used for large-scale production of electricity without greenhouse gas generation, advanced medical treatments, food sterilization, and sophisticated measurement technologies. In order to benefit from these uses, nuclear materials packaging must be developed, manufactured, inspected, and maintained for storage, transfer, transport, and disposal that protect the public and the environment. The U.S. Department of Energy (DOE) Packaging Certification Program (PCP) originally offered the following ten, one-day to ten-day training courses. 1. Methods for Reviewing Safety Analysis Reports for Packages and Performing Confirmatory Analysis; 2. Welding Criteria for Shipping Containers; 3. Structural Analysis and Acceptance Criteria; 4. Quality Assurance (QA) for Radioactive Material Packaging Training; 5. Application of the American Society of Mechanical Engineers (ASME) Code to Radioactive Material Transportation Packaging; 6. ARG-US Radio Frequency Identification (RFID) System for Management of Nuclear and Radioactive Materials; 7. Packaging Operations and Maintenance; 8. Packaging Leakage Test Training; 9. Management of SARP Preparation; 10. Transport Security of Radioactive Material. These courses were designed in accordance with DOE standards. They were taught at a rigorous graduate-university level by experienced technical staff who are coordinated by Argonne, Lawrence Livermore, and Savannah River National Laboratories. The courses were taught at these Laboratories or nearby locations. The PCP courses are attended by personnel from government agencies, national laboratories, and the nuclear industry, from both inside and outside the United States. Students in these classes have generally earned undergraduate or higher degrees. These courses are designed to assist student with specialized job duties, and may provide them a competitive advantage when seeking employment or promotion in the nuclear packaging and other related fields. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

High Temperature Sealants and Gaskets in Nuclear Applications

Lawrence Livermore National Laboratory (LLNL) has been conducting research to develop and deploy advances in ceramic filter technology to benefit nuclear facilities. There is a need for better materials for High Efficiency Particulate Air (HEPA) filters, especially those able to withstand moisture and higher temperatures as experienced in fire conditions. One of the next steps in this process is the improvement of the system sealants and gaskets. The LLNL team has reached out to the University of Texas at El Paso (UTEP) for assistance with the task of identifying potential materials as well as conducting experiments utilizing the testing units developed by LLNL's collaboration with California Polytechnic State University (Cal Poly). Current commercial filters use fluid seals and urethane gasket seals. Although these seals are sufficient for the minimum requirements of existing nuclear facilities, they will not be sufficient as improvements in ceramic filters and pre-filters reach their full potential. Currently, HEPA filters can only survive up to 1000 deg. F (exhaust air) and 500 deg. F (supply air) for less than 2 hours. Our research at the University of Texas at El Paso reviewed hundreds of sealant and gasket materials to identify improved sealants and gaskets that can be used in existing filters as well as ceramic HEPA filters and pre-filters. Results of the research will be supplied to support the relevant code, American Society of Mechanical Engineers (ASME) AG-1, in particular Section FO. (authors)

42 ENGINEERING↗
DOE OSTI2017

Uniaxial Ratcheting and Fatigue Failure of SA508-3 at Elevated Temperature

Ratcheting, describing the accumulation of cyclic plastic deformation under asymmetric cyclic stressing conditions, is a significant factor in the evaluation of fatigue life. SA508-3 ferritic stainless steel is a key material in the first loop of pressurized water reactor, often subjected to cyclic thermal and mechanical loadings with a temperature range of 25 deg. C to 350 deg. C. Therefore, the ratcheting behaviors of SA508-3 steel in service attract enough attentions. Both the American ASME and French RCC-M codes give specific descriptions about the assessment of the ratcheting behavior. However, it is difficult to predict the ratcheting strain since it is a secondary deformation superposed to the primary stress-strain response. Recent decades, the ratcheting behaviors of metals have been extensively studied experimentally and theoretically. Many of these researches mainly focused on the experiment observation of ratcheting behaviors within hundreds number of cycles and constitutive modeling without considering damage. A damage-coupled viscoplastic model was proposed by Kang et al. to simulate the whole-life ratcheting and predict the fatigue life of the tempered 42CrMo steel successfully, and thus it is worth to investigate the ratcheting-fatigue interaction for SA508-3 steel.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Fabrication and Testing of DOE Standard Canister Closure Leak Test Assembly

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.

42 ENGINEERING↗
DOE OSTI2022

Alloy 709 Advanced Austenitic Stainless Steel

Provide material solutions to enabling design, construction, licensing and operation of advanced reactors Including Fast Reactors, Gas-cooled Reactors and Molten Salt Reactors (solid or liquid fuel) Could be of modular or micro designs, and from hundreds of MWe to kWe Provide technical bases needed for design and licensing of advanced reactor components Develop & validate improved high temperature design methodology Provide qualification data (to NQA-1 or equivalent) on structural materials Utilize consensus standards when appropriate (e.g., ASME, ASTM, etc.) Target resolution of issues needed for near to mid-term deployment of advanced reactors

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗