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At least 487 records · Page 27

Bifurcation Buckling Analysis and Non-Linear Collapse Analysis of Teardrop Shaped Vacuum Chamber

The Spallation Neutron Source (SNS) accelerator is being upgraded to increase the beam power from 1.4MW at 1GeV to 2.8MW at 1.3GeV. The currents in the middle two injection chicane magnets cannot simply be scaled up to accommodate the increased injection energy of 1.3GeV due to potential excessive H- stripping; the magnets must be replaced with longer, lower-field magnets and the associated vacuum chambers need to be redesigned. A new teardrop-shaped vacuum chamber was initially designed to accommodate the new magnets and the updated beam paths and instrumentation. This paper focuses on the structural stability study of the teardrop shape vacuum chamber based on buckling analysis. Protection against collapse from buckling according to the ASME BPVC requirement has been evaluated in depth. First, a Type-1 bifurcation buckling analysis using a linear eigenvalue solution to determine the critical load factor was performed. Subsequently, a Type-3 nonlinear collapse analysis was conducted using the static Riks method with elastic-plastic material properties and imperfections explicitly considered in the model geometry. The critical buckling load for the teardrop shape vacuum chamber was confidently estimated based upon this two-stage approach.

Jiang, Hao↗

STRUCTURAL MODELING TO SUPPORT POST-YIELD ACCEPTANCE CRITERIA FOR SPENT NUCLEAR FUEL CLADDING

Spent nuclear fuel (SNF) is evaluated for structural failure during storage and transportation scenarios. The U.S. Department of Energy’s Spent Fuel and Waste Science and Technology (SFWST) program has sponsored significant research in quantifying mechanical loads on SNF during storage and transportation scenarios using experimental and modeling methods. The SFWST program has also performed significant research on measuring the mechanical behavior of irradiated SNF as defueled cladding segments and cladding with fuel pellets to measure composite behavior. This paper considers some of the key material data from the Sibling Pin testing and uses structural modeling and analysis methods that have been informed by testing to consider post-yield acceptance criteria for SNF cladding structural analysis. Test data published by Oak Ridge National Laboratory (ORNL) and Pacific Northwest National Laboratory (PNNL) are the foundation for informing the material behavior of the models developed in this study. In particular, four-point bend (4PB) tests of fueled and defueled cladding segments provide significant information about the bending failure mode of SNF. ORNL’s 4PB test data is on fueled cladding segments, so the composite behavior of SNF is demonstrated. This paper describes PNNL’s coincident beam model that was developed to approximate the composite behavior of SNF. This paper also presents PNNL’s structural dynamic finite element models of a cask tip-over scenario, which is predicted to cause the strongest mechanical loads on SNF of all postulated storage and transportation scenarios. SNF bending loads predicted in the cask tip-over scenario and cladding acceptance criteria beyond yield are considered, with justification based on the Sibling Pin test data. ASME Boiler and Pressure Vessel code stress intensity limits are also considered. The ultimate goal of this work is to aid in the justification of structural acceptance criteria for SNF cladding beyond the cladding’s irradiated yield strength for use in structural analysis of all storage and transportation scenarios.

Klymyshyn, Nicholas A.↗

Influence of Proton Activity Gaps between Electrodes on Open-Circuit Potential of H 2 /H 2 and H 2 /Air Cells

Polymer -electrolyte-fuel-cell open-circuit voltages (OCVs) are exactly defined by equation (1), where cathode and anode proton activities [(aH+)cathode and (aH+)anode, respectively] usually are identical, so the third term in the right-hand side of equation (2) is ignored. OCV=E0+RT/2F*ln(a1/2 O2*(a2 H+)cathode/aH2O)-RT/2F*ln((a2 H+)anode/aH2) (1) =E0+RT/2F*ln(a1/2 O2*aH2/aH2O)+RT/2F*ln((a2 H+)cathode/(a2 H+)anode) (2) Water vapor pressure is a colligative property that fundamentally correlates to electrolyte concentrations in aqueous solutions. Proton activity is a function of acid concentration, such as pH, when electrolytes are acids. In polymer-electrolyte membranes, water vapor pressure and acid concentration are understood as relative humidity (RH) and water uptake (λ), respectively, where λ represents number of water molecules per sulfonic acid molecule. Several investigations have reported the relationship between RH and λ, meaning that proton activities and associated water uptakes are intimately related to RH. In actual fuel cell operation, cathode RH is determined by ambient-atmosphere and/or humidifier RH(s), and anode RH depends on hydrogen-circulator RH. Therefore, RH is not always identical at both electrodes, and the difference between electrode RHs is considerable during dry operation of polymer electrolyte fuel cells. Therefore, the third term in the right-hand side of equation (2) may be significant for dry operation. We measured OCVs when hydrogen was supplied to both electrodes at 80°C. One electrode (A) was fixed at 30% RH, while RH at the other electrode (B) was varied (0, 5, 10, 20, and 30%). Measured OCVs varied from 0 to 75 mV. For fuel cell tests, electrode A was supplied with hydrogen at 30% RH; electrode B, oxygen at 0, 5, 10, 20, and 30% RH. OCVs deviated from that measured when RH at electrode B was 30%, increasing from 0 to 60 mV with decreasing RH at electrode B. Results are also shown in Figure 1. Proton activities of both electrodes were thermodynamically calculated. The Gibbs–Duhem relation was applied to obtain molar Gibbs free energies of water and sulfonic acid, and proton activity coefficient was calculated using the Gibbs free energy of sulfonic acid and the relationship between RH and λ1–4, assuming that protons and sulfonic anions show identical ionic-activity coefficients. OCVs were estimated using the third term in the right-hand side of equation (2). Results are shown in Figure 1. Fuel-cell current–voltage performance was poor when RHs at the anode and cathode were 30 and 20%, respectively. To determine kinetic current, we measured the oxygen-reduction reaction (ORR) using a rotating-disk electrode (RDE) in concentrated-acid aqueous solutions, which modeled catalyst-layer ionomers. Kinetic currents decreased with acid concentrations. References T. A. Zawodzinski, Jr., C. Derouin, S. Radzinski, R. J. Sherman, V. T. Smith, T. E. Springer and S. Gottesfeld , J. Electrochem. Soc., 140,1041 (1993) P. K. Das and A. Z. Weber, Proceedings of the ASME 2013 11th Fuel Cell Science, Engineering and Technology Conference, Fuel Cell 18010 (2013) V. A. Sethuraman, J. W. Weidner, A. T. Haug, S. Motupally,b and L. V. Protsailo, J. Electrochem. Soc., 155, B50 (2008) A. Kusoglu and A. Z. Weber, Chem. Rev., 117, 987 (2017) Figure 1

Yoshida, Toshihiko↗

BlendPATH (Blending Pipeline Analysis Tool for Hydrogen) [SWR-24-10]

BlendPATH provides case-by-case techno-economic analysis for potential projects where hydrogen is blended into a natural gas transmission pipeline. BlendPATH estimates 1) the transmission pipeline modifications and operating conditions necessary to blend hydrogen to a user specified volume faction of hydrogen in pipeline gas and 2) the incremental capital and operating expenses to prepare said transmission pipeline for hydrogen blending. BlendPATH is developed in Python and requires SAInt, a underlying commercial natural gas pipeline network modeling software, to run. Pipeline modification and operation condition estimation is guided by ASME B31.12. The intent of this software is to target application for projects in the initial project assessment stage and provide the user with the capability of assess promising opportunities before the use proceeds with further detailed pipeline evaluation based on a probable economic outcome.

Kee, Jamie↗

Compilation of Published Test Data for Cylindrical and Spherical Vessels under External Pressure

This dataset is a compilation of the results of previously-published failure tests of cylindrical and spherical pressure vessels, pipes, shells, and caps under external pressure. This data has been used for the article 'Simplified Formulas for External Pressure Design' by W. Kirkland and C. Bett, presented at the ASME Pressure Vessels and Piping Conference 2022. Paper number PVP2022-78354.

42 ENGINEERING↗

Acquired experience resulting from transforming a chemical installation into a nuclear one

ICIT-Valcea has developed an experimental pilot-scale installation for tritium and deuterium separation. The main objective of this pilot was to demonstrate the water detritiation technology and to transfer this technology to the CANDU reactors of the Cernavoda nuclear power plant. The pilot-scale installation was initiated in 1992. The initial design and construction were performed similarly to chemical plants as the separation of isotopes was focused on only hydrogen and deuterium to assess feasibility. In a second phase we have begun to transform it into a nuclear facility with the aim of separating tritium. Moving to tritium separation has imposed a lot of changes. Changes consisted mainly of: -) re-design of the technological systems for nuclear material processing, applying specific codes and standards (ASME, Romanian nuclear specific pressure boundary prescriptions for code classification); -) design and implementation of new systems, classified as safety systems; -) re-design and implementation of command and control systems, complying with the requirements of reliability and maintenance required for the project promoted; -) revaluation of auxiliary systems (utilities, power supply); -) implementing radiation protection systems, including secondary barriers; -) implementing and maintaining environment operational program specific to the new nuclear plant; -) developing and conducting safety analyzes; and -) the production of specific documentation to obtain the necessary permits for construction, commissioning and operation of the plant.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Optimization of a cyclone using MFIX and Nodeworks

Video depicting the optimization process of a cyclone on NETL's chemical looping reactor (CLR) using MFIX and Nodeworks. MFIX is used to model the cyclone using PIC. Nodeworks is then used to generate proposed geometry changes using a Latin hypercube. Each design is simulated, with an objective value being computed based on the cyclone efficiency and pressure drop. A Gaussian Process surrogate model is then constructed from the objective values. This surrogate model is then used by a differential evolution optimization algorithm to identify the optimal cyclone design. Details published here: Weber, J., Fullmer, W., Gel, A., and Musser, J. (February 4, 2020). "Optimization of a Cyclone Using Multiphase Flow Computational Fluid Dynamics." ASME. J. Fluids Eng. March 2020; 142(3): 031111. https://doi.org/10.1115/1.4045952 OSTI: https://www.osti.gov/pages/servlets/purl/1763893

cyclone↗

Pressure Spike in The LBNF Absorber Core s Gun Drilled Cooling Channel from an Accident Beam Pulse

The LBNF Absorber consists of thirteen 6061-T6 aluminum core blocks. The core blocks are water cooled with de-ionized (DI) water which becomes radioactive during beam operations. The cooling water flows through gun-drilled channels in the core blocks. The cooling water is supplied by the LBNF Absorber Radioactive Water (RAW) cooling system which is designed as per ASME B31.3 Normal Fluid Service [1]. An uninhibited beam accident pulse striking the water channels was identified as a credible accident scenario. In this study, it is assumed that the beam pulse hits the Absorber directly without interacting with any of the other upstream beamline components. The beam parameters used for the LBNF beam are 120 GeV, 2.4 MW with a 1.2 s cycle time. The accident pulse lasts for 10 μs. The maximum energy is deposited in the 3rd aluminum core block. For the sake of simplicity, it is assumed that the accident pulse strikes the 1 in. ID water channel directly. The analysis here simulates the pressure rise in the water during and after the beam pulse and its effects on the aluminum piping components that deliver water to the core blocks. The weld strengths as determined by the Load and Resistance Factor Design (LRDF) and the Allowable Strength Design (ASD) are compared to the forces generated in the weld owing to the pressure spike. A transient structural analysis was used to determine the equivalent membrane, peak, and bending stresses and they were com-pared to allowable limits.

43 PARTICLE ACCELERATORS↗

Optimization of a Welding Procedure for Making Critical Aluminum Welds on the LBNF Absorber Core Block

The LBNF Absorber consists of thirteen 6061-T6 aluminum core blocks. The core blocks are water cooled with de-ionized (DI) water which becomes radioactive during beam operations. The cooling water flows through gun-drilled channels in the core blocks. A weld quality optimization was performed to produce National Aeronautical Standard 1514 Class I quality welds on the aluminum core blocks. This was not successful in all cases. An existing Gas Tungsten Arc Welding Procedure Specification was fine tuned to minimize, in most cases, and eliminate detect-able tungsten inclusions in the welds. All the weld coupons, however passed welding inspection as per the piping code: ASME B31.3 Normal Fluid Service. Tungsten electrode diameter, type, and manufacturer were varied. Some of the samples were pre-heated and others were not. It was observed that larger diameter electrodes, 5/32 in., with pre-heated joints resulted in welds with the least number of tungsten inclusions. It is hypothesized that thinner electrodes breakdown easily and get lodged into the weld pool during the welding process. This breakdown is further enhanced by the large temperature differential between the un-preheated sample and the hot electrode.

43 PARTICLE ACCELERATORS↗

AGC-2 Disassembly Report

The Next Generation Nuclear Plant (NGNP) Graphite Research and Development (R&D) Program is currently measuring irradiated material properties for predicting the behavior and operating performance of new nuclear graphite grades available for use within the cores of new very high temperature reactor designs. The Advanced Graphite Creep (AGC) experiment, consisting of six irradiation capsules, will generate irradiated graphite performance data for NGNP reactor operating conditions. The AGC experiment is designed to determine the changes to specific material properties such as thermal diffusivity, thermal expansion, elastic modulus, mechanical strength, irradiation induced dimensional change rate, and irradiation creep for a wide variety of nuclear grade graphite types over a range of high temperature, and moderate doses. A series of six capsules containing graphite test specimens will be used to expose graphite test samples to a dose range from 1 to 7 dpa at three different temperatures (600, 900, and 1200°C) as described in the Graphite Technology Development Plan. Since irradiation induced creep within graphite components is considered critical to determining the operational life of the graphite core, some of the samples will also be exposed to an applied load to determine the creep rate for each graphite type under both temperature and neutron flux. All six AGC capsules in the experiment will be irradiated in the Advanced Test Reactor (ATR). AGC-1 and AGC-2 will be irradiated in the south flux trap and AGC-3–AGC-6 will be irradiated in the east flux trap. The change in flux traps is due to NGNP irradiation priorities requiring the AGC experiment to be moved to accommodate Fuel irradiation experiments. After irradiation, all six AGC capsules will be cooled in the ATR Canal, sized for shipment, and shipped to the Materials and Fuels Complex (MFC) where the capsule will be disassembled in the Hot Fuel Examination Facility (HFEF). During disassembly, the metallic capsule will be machined open and the individual samples removed from the interior graphite body containing the samples. Samples removed from the capsule will be loaded in a shipping drum and shipped to the Idaho National Laboratory (INL) Research Center (IRC) for initial post-irradiation examination (PIE) and storage for any future testing at the newly completed Carbon Characterization Laboratory (CCL). All work was performed under an ASME NQA-1-2008;1a-2009 compliant quality assurance program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Creep and Creep-Fatigue of Alloy 617 Weldments

Alloy 617 is the primary candidate material for the heat exchanger of a very high temperature gas cooled reactor intended to operate up to 950°C. While this alloy is currently qualified in the ASME Boiler and Pressure Vessel Code for non-nuclear construction, it is not currently allowed for use in nuclear designs. A draft Code Case to qualify Alloy 617 for nuclear pressure boundary applications was submitted in 1992, but was withdrawn prior to approval. Prior to withdrawal of the draft, comments were received indicating that there was insufficient knowledge of the creep and creep-fatigue behavior of Alloy 617 welds. In this report the results of recent experiments and analysis of the creep-rupture behavior of Alloy 617 welds prepared using the gas tungsten arc process with Alloy 617 filler wire. Low cycle fatigue and creep-fatigue properties of weldments are also discussed. The experiments cover a range of temperatures from 750 to 1000°C to support development of a new Code Case to qualify the material for elevated temperature nuclear design. Properties of the welded material are compared to results of extensive characterization of solution annealed plate base metal.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Environmental Virology Workshop Summary, Tucson, Arizona, Jan 7-12, 2013

Full Text of the report: A total of 66 researchers participated in this workshop, including 44 attendees, 3 program officers from private and federal funding agencies, and 19 workshop teachers. The workshop was incredibly productive and focused on identifying knowledge-gaps critical for predictive modeling, and developing the framework (experimental, informatic, theoretical) needed to obtain the data. All attendees developed a strong foundation in cutting-edge methods and a network of researchers that are now aiding in advancing environmental virology research. To more broadly reach Environmental Virologists, a subset of the attendees since proposed and ran a viromics workshop at the American Society of Microbiology meeting in 2014 in Boston, MA where the workshop sold-out. The workshop proposal was accepted again by ASM and is scheduled to occur at the New Orleans meeting in May, 2015. Additionally, PI Sullivan is co-convening a ''Viromics: Tools and Concepts'' session at the FEMS meeting in the Netherlands in June 2015 to continue getting the word out about Environmental Virology. A second formal Environmental Virology Workshop is being planned to occur in Scotland in summer 2016, likely held jointly with the Aquatic Virology Workshop. I wish to thank DOE for their critical support for this workshop which has helped galvanize the field.

54 ENVIRONMENTAL SCIENCES↗

Comprehensive Report For Proposed Elevated Temperature Elastic Perfectly Plastic (EPP) Code Cases Representative Example Problems

Background: The current rules in the nuclear section of the ASME Boiler and Pressure Vessel (B&PV) Code , Section III, Subsection NH for the evaluation of strain limits and creep-fatigue damage using simplified methods based on elastic analysis have been deemed inappropriate for Alloy 617 at temperatures above 1200F (650C)1. To address this issue, proposed code rules have been developed which are based on the use of elastic-perfectly plastic (E-PP) analysis methods and which are expected to be applicable to very high temperatures. The proposed rules for strain limits and creep-fatigue evaluation were initially documented in the technical literature 2, 3, and have been recently revised to incorporate comments and simplify their application. The revised code cases have been developed. Task Objectives: The goal of the Sample Problem task is to exercise these code cases through example problems to demonstrate their feasibility and, also, to identify potential corrections and improvements should problems be encountered. This will provide input to the development of technical background documents for consideration by the applicable B&PV committees considering these code cases for approval. This task has been performed by Hollinger and Pease of Becht Engineering Co., Inc., Nuclear Services Division and a report detailing the results of the E-PP analyses conducted on example problems per the procedures of the E-PP strain limits and creep-fatigue draft code cases is enclosed as Enclosure 1. Conclusions: The feasibility of the application of the E-PP code cases has been demonstrated through example problems that consist of realistic geometry (a nozzle attached to a semi-hemispheric shell with a circumferential weld) and load (pressure; pipe reaction load applied at the end of the nozzle, including axial and shear forces, bending and torsional moments; through-wall transient temperature gradient) and design and operating conditions (Levels A, B and C).

36 MATERIALS SCIENCE↗

Define and Quantify the Physics of Air Flow, Pressure Drop and Aerosol Collection in Nuclear Grade HEPA Filters

Objective: Develop a set of peer-review and verified analytical methods to adjust HEPA filter performance to different flow rates, temperatures and altitudes. Experimental testing will measure HEPA filter flow rate, pressure drop and efficiency to verify the analytical approach. Nuclear facilities utilize HEPA (High Efficiency Particulate Air) filters to purify air flow for workspace ventilation. However, the ASME AG-1 technical standard (Code on Nuclear Air and Gas Treatment) does not adequately describe air flow measurement units for HEPA filter systems. Specifically, the AG-1 standard does not differentiate between volumetric air flow in ACFM (actual cubic feet per minute)compared to mass flow measured in SCFM (standard cubic feet per minute). More importantly, the AG-1 standard has an overall deficiency for using HEPA filter devices at different air flow rates, temperatures, and altitudes. Technical Approach: The collection efficiency and pressure drops of 18 different HEPA filters will be measured over a range of flow rates, temperatures and altitudes. The experimental results will be compared to analytical scoping calculations. Three manufacturers have allocated six HEPA filters each for this effort. The 18 filters will be tested at two different flow rates, two different temperatures and two different altitudes. The 36 total tests will be conducted at two different facilities: the ATI Test facilities (Baltimore MD) and the Los Alamos National Laboratory (Los Alamos NM). The Radiation Protection RP-SVS group at Los Alamos has an aerosol wind tunnel that was originally designed to evaluate small air samplers. In 2010, modifications were started to convert the wind tunnel for HEPA filter testing. (Extensive changes were necessary for the required aerosol generators, HEPA test fixtures, temperature control devices and measurement capabilities.) To this date, none of these modification activities have been funded through a specific DOE or NNSA program. This is expected to require six months of time, after receipt of funding. Benefits: US DOE facilities that use HEPA filters will benefit from access to the new operational measurement methods. Uncertainty and guesswork will be removed from HEPA filter operations.

61 RADIATION PROTECTION AND DOSIMETRY↗

AGR-5/6/7 LEUCO Kernel Fabrication Readiness Review

In preparation for forming low-enriched uranium carbide/oxide (LEUCO) fuel kernels for the Advanced Gas Reactor (AGR) fuel development and qualification program, Idaho National Laboratory conducted an operational readiness review of the Babcock & Wilcox Nuclear Operations Group – Lynchburg (B&W NOG-L) procedures, processes, and equipment from January 14 – January 16, 2015. The readiness review focused on requirements taken from the American Society Mechanical Engineers (ASME) Nuclear Quality Assurance Standard (NQA-1-2008, 1a-2009), a recent occurrence at the B&W NOG-L facility related to preparation of acid-deficient uranyl nitrate solution (ADUN), and a relook at concerns noted in a previous review. Topic areas open for the review were communicated to B&W NOG-L in advance of the on-site visit to facilitate the collection of objective evidences attesting to the state of readiness.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

System-Level Heat Transfer Analysis, Thermal- Mechanical Cyclic Stress Analysis, and Environmental Fatigue Modeling of a Two-Loop Pressurized Water Reactor. A Preliminary Study

This report provides an update on an assessment of environmentally assisted fatigue for light water reactor components under extended service conditions. This report is a deliverable in April 2015 under the work package for environmentally assisted fatigue under DOE's Light Water Reactor Sustainability program. In this report, updates are discussed related to a system level preliminary finite element model of a two-loop pressurized water reactor (PWR). Based on this model, system-level heat transfer analysis and subsequent thermal-mechanical stress analysis were performed for typical design-basis thermal-mechanical fatigue cycles. The in-air fatigue lives of components, such as the hot and cold legs, were estimated on the basis of stress analysis results, ASME in-air fatigue life estimation criteria, and fatigue design curves. Furthermore, environmental correction factors and associated PWR environment fatigue lives for the hot and cold legs were estimated by using estimated stress and strain histories and the approach described in NUREG-6909. The discussed models and results are very preliminary. Further advancement of the discussed model is required for more accurate life prediction of reactor components. This report only presents the work related to finite element modelling activities. However, in between multiple tensile and fatigue tests were conducted. The related experimental results will be presented in the year-end report.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Oil-free centrifugal hydrogen compression technology demonstration

One of the key elements in realizing a mature market for hydrogen vehicles is the deployment of a safe and efficient hydrogen production and delivery infrastructure on a scale that can compete economically with current fuels. The challenge, however, is that hydrogen, being the lightest and smallest of gases with a lower viscosity and density than natural gas, readily migrates through small spaces and is difficult to compresses efficiently. While efficient and cost effective compression technology is crucial to effective pipeline delivery of hydrogen, the compression methods used currently rely on oil lubricated positive displacement (PD) machines. PD compression technology is very costly, has poor reliability and durability, especially for components subjected to wear (e.g., valves, rider bands and piston rings) and contaminates hydrogen with lubricating fluid. Even so called “oil-free” machines use oil lubricants that migrate into and contaminate the gas path. Due to the poor reliability of PD compressors, current hydrogen producers often install duplicate units in order to maintain on-line times of 98-99%. Such machine redundancy adds substantially to system capital costs. As such, DOE deemed that low capital cost, reliable, efficient and oil-free advanced compressor technologies are needed. MiTi’s solution is a completely oil-free, multi-stage, high-speed, centrifugal compressor designed for flow capacity of 500,000 kg/day with a discharge pressure of 1200 psig. The design employs oil-free compliant foil bearings and seals to allow for very high operating speeds, totally contamination free operation, long life and reliability. This design meets the DOE’s performance targets and achieves an extremely aggressive, specific power metric of 0.48 kW-hr/kg and provides significant improvements in reliability/durability, energy efficiency, sealing and freedom from contamination. The multi-stage compressor system concept has been validated through full scale performance testing of a single stage with helium similitude gas at full speed in accordance with ASME PTC-10. The experimental results indicated that aerodynamic performance, with respect to compressor discharge pressure, flow, power and efficiency exceeded theoretical prediction. Dynamic testing of a simulated multistage centrifugal compressor was also completed under a parallel program to validate the integrity and viability of the system concept. The results give strong confidence in the feasibility of the multi-stage design for use in hydrogen gas transportation and delivery from production locations to point of use.

08 HYDROGEN↗

Target Housing Material Options

With gas cooling, heat transfer coefficients are low compared to water. The benefit of gas from a heat transfer point of view is that there is really no upper temperature limit for the coolant, as compared to water, which is limited ultimately by the critical point, and in practice the critical heat flux. In our case with parallel flow channels, water is limited to even lower operating limits by nucleate boiling. So gas can get as hot as the containment material will allow, but to get the density and heat transfer up to something reasonable, we must also increase pressure, thus increasing stress on the containment, namely the front and back faces. We are designing to ASME BPVC, which, for most materials allows a maximum stress of UTS/3. So we want the highest possible UTS. For reference, the front face stress in the 12 mm target at 300 psi was about 90 MPa. The inconel 718 allowable stress at 900°C is 1/3 of 517 or 172 MPa. So we are in a very safe place, but the uTS is dropping rapidly with temperature above 900°C. As we increase target diameter, the challenge will be to keep the stress down. We are probably looking at keeping the allowable at or above the present value, and at as high a temperature as possible.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗