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All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

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

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

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Drop Analysis of a Department of Energy Standard Canister Containing Fort Saint Vrain SNF

DOE manages over 300 types of SNF, many of which are located at the 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 (RRDS) is an SNF management concept in which SNF is packaged into dry, sealed canisters that are then placed in on-site 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 RRDS program at the INL site. One critical aspect of RRDS is the ability to certify the DOE Standard Canister and its associated transportation package in accordance with 10 CFR 71 for offsite transportation. Depending on the SNF type and transportation strategy, DOE Standard Canisters may be required to maintain structural integrity under hypothetical accident scenarios (e.g., drop events). The DOE Standard Canisters have been tested and analyzed under various SNF loading configurations and accident drop events in support of the Idaho Spent Fuel Facility and other DOE programs; however, no analysis has yet been completed in support of the recently initiated Road-Ready Demonstration. This paper presents preliminary results from a finite element analysis of the Ø45.7 cm × 4.6 m (Ø18 in. × 15 ft) DOE Standard Canister under the 9 m drop at 80 degrees off-vertical drop scenario considered in previous INL tests and analyses. It considers the Fort St. Vrain spent nuclear fuel loading configuration proposed for the Road-Ready Demonstration, uses updated material properties, and applies the strain-based acceptance criteria established in ASME Boiler and Pressure Vessel Code’s Section III, Division 3 rules for storage and transportation spent nuclear fuel containments. This updated analysis is compared to previous DOE Standard Canister drop analyses. Preliminary results from the updated analysis show that certain regions of the containment exceed the allowable limits during the accidental drop event. However, these regions are limited to components performing a non-structural function. While further work on this analysis will be pursued, this analysis serves as the foundation for formal calculations used to support applicable certification efforts of the RRDS system at INL.

42 ENGINEERING↗
DOE OSTI2024

Development of creep resistant ferritic-martensitic steels for operation at 650°C

Creep resistant ferritic-martensitic steels capable of operating at 650°C have long been under development for electricity generation. This presentation will showcase the design and development of alloys CPJ-7 and JMP with approximately 420,000 h of in-house creep testing, the equivalent of ~50 years of cumulative creep tests. Approximately fifty different alloy compositions were manufactured as part of an ICME approach. CPJ-7 present superior creep properties compared to COST alloys for steam turbine and P91/92 for boiler applications. The prolonged creep life was attributed to slowing down the destabilization of fine precipitates at 650°C. The JMP steels were designed with higher Co for solid solution strengthening and increased W for matrix strength and stability. The creep life compared to CPJ-7 increased between 118 and 150% at 650°C depending on the stress. On a Larson-Miller plot, the performance of the JMP steels surpasses that of MARBN steel. Composition-microstructure-properties relationships will be discussed.

Detrois, Martin↗
DOE OSTI2024

Technology Feasibility of High-Temperature Heat Pumps for Industrial Drying Process in the US

High-temperature heat pumps are a prospective technology for electrifying and decarbonizing industrial drying processes. This study investigates the technological potential of HTHPs in replacing fossil-fuel burners or steam boilers in industrial spray dryers. The performance of HTHPs with three configurations was investigated, including high high-temperature heat pump boosted with an electric air heater, a steam-generating high-temperature heat pump with multi-stage mechanical vapor compressions, and transcritical high-temperature heat pumps. R1336mzz(Z) is used as the refrigerant in both subcritical and transcritical high-temperature heat pumps, and R717 is used as the working fluid in the mechanical vapor recompression and the heat transfer fluids. Thermodynamic models are built to evaluate the energy efficiency of HTHPs. A case study was carried out on an industrial spray dryer integrated with a waste heat recuperator, where the HTHPs are used to recover precooled air with a dew point of 40℃ and supply heat for drying air at 200℃. For a high-temperature heat pump boosted with an electric air heater, a higher system-level COP is achieved with a high supply temperature from high-temperature heat pump. For a high-temperature heat pump with mechanical vapor recompressions, larger temperature drops in the flashing tank provide higher heat capacities but almost constant values of the coefficient of performance. For a transcritical high-temperature heat pump, an optimum discharge pressure exists for the temperature profiles within the gas cooler and the coefficient of performance. The transcritical high-temperature heat pump provides the best energy efficiency among the three configurations. The theoretical results prove the technical feasibility of high-temperature heat pumps in industrial air-drying applications. .

Wang, Pengtao↗
DOE OSTI2024

ASME Design Code Rule Changes for Nuclear Graphite

The American Society of Mechanical Engineers Boiler Pressure and Vessel Code (ASME BPVC) Section III, Division 5, Article HHA-3000 outlines graphite core component and graphite core assembly design guidelines. Graphite core components are defined as ?components manufactured from graphite that are installed to form a graphite core assembly within the reactor pressure vessel of a high temperature, graphite moderated fission reactor.? (p. 413) Graphites? inherent defect distributions do not allow for deterministic material reliability. Rather, graphite has variable strength distributions which change by grade. Article HHA-3000 outlines two semi-probabilistic methods, the full and simplified assessments, which set design load limit targets for each of three component structural reliability classes. The Design Task Group was officially recognized as a specialized task group within ASME November of 2023, though we?ve been collaborating since 2022. The purpose of the Design Task Group is to correct, clarify, and make HHA-3000 function as intended. The Design Task Group will sunset once we?ve achieved our objectives. The Design Task Group was specifically told to not write new Code. While there may be more precise and more accurate methods to determine reliability targets, the current methods are conservative, relatively simple to implement, and have thus far been considered satisfactory for setting design reliability targets. Much of the ground-work to write proposal files and background documents for records to make the changes needed to achieve our objective have been completed. The Design Task Group has documented much of their work through papers, presentations, and memorandums. Three memorandums in which INL team members had substantial contributions are found in the Appendices: FEA Modeling for the Baseline Program, Evaluating the Effects on Margin of Updating the Threshold and Shape Parameters in the Full Assessment, and Interpretations of the Full and Simplified Assessments in ASME BPVC. Most of the on-going work to achieve the Design Task Group?s objective will be addressing comments on existing records and moving records through the balloting process. The Design Task Group met bi-weekly mostly through the end of FY2023. Since February 2024, the Design Task Group has mostly been completed with solving and documenting the technical issues associated with the assessments. Unless new tasks are identified, the remaining work of the Design Task Group will be political and editorial.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2023

Assessment of UTP A 2133 Mn as a Matching Filler Metal for Alloy 800H in Section III, Division 5 Applications

Six alloys are qualified in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC) for elevated-temperature nuclear components. Due to reactor design alloy restrictions and motivation to improve operational efficiency and plant lifetimes, it is necessary to maximize the usefulness of qualified alloys. One of these alloys is Alloy 800H, which is qualified for a maximum temperature of 760°C and a maximum 300,000-hour service life. However, the welding filler metals qualified to join Alloy 800H have stress rupture factors that reduce the allowable weldment strength by factors as low as 0.59 times the Alloy 800H base metal. In an effort to improve the weldment creep-rupture performance, or to increase the stress reduction factor, non-code qualified filler metals are under investigation by the Department of Energy?s Advanced Reactor Technologies program. For this investigation, UTP A 2133 Mn filler metal (Fe-Cr-Ni-Mn-Nb) was used to join 0.5 in. thick Alloy 800H plate to demonstrate passing an ASME BPVC Section IX weld qualification. The multi-pass, pulsed GTAW process showed successful weld qualification results, and additional property measurements were conducted to compare previous analyses of Alloy 82 and Alloy 617 filler metals. Future work will evaluate the cross-weld creep-rupture performance of the UTP A 2133 Mn in comparison to Alloy 82 and Alloy 617 filler metals.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

FY24 progress report on A709 mechanical properties data development and A709 thermal aging status

The report provides the status of the creep, fatigue, and creep-fatigue testing to date conducted at Idaho National Laboratory to generate the data package. This data package evaluates the material performance from three commercial heats to support the Alloy 709 qualification in American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section III, Division 5. First procured heat was manufactured by G.O. Carlson with heat number 58776. Second and third heats were fabricated by ATI Specialty Rolled Products with heat numbers 529900 and 530843, respectively. A series of creep and cyclic specimens were fabricated from these three heats and tests were performed. A list of finished, ongoing, and planned test are presented for creep, fatigue, and creep-fatigue tests. Cyclic properties of three commercial heats are compared.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Creep-Fatigue Properties of Additional 316H PM-HIP Materials Fabricated from Different Powder Compositions and Processing Routes

The process of powder metallurgy (PM) hot isostatic pressing (HIP) works by consolidating powdered materials at relatively high temperature and pressure to form near-net-shaped components. Ideally, PM-HIP production methods can reduce component lead time and improve designs for high-temperature reactors and/or microreactors. To introduce PM-HIP into Section III, Division 5 of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, it is necessary to show adequate material properties regarding creep, high-temperature low-cycle fatigue, and creep fatigue. However, prior work has shown that the creep-fatigue cycles to failure for PM-HIP 316H stainless steel are greatly reduced compared to the conventional, wrought product. This work continued creep-fatigue analysis on a 316H stainless steel with lower oxygen and nitrogen contents and at different HIP parameters than previously analyzed. The objective was to better understand what is causing the reduced PM-HIP 316H performance so improvements can be made PM-HIP 316H creep-fatigue lifetimes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Performance Validation of a Thermally Integrated 50 kW High Temperature Electrolyzer System

In the proposed project, INL and OxEon seek to improve the value proposition of hydrogen production by integrating reversible fuel cell operations at relatively small scale for distributed energy applications. This goal will be accomplished by converting a 50 kW solid oxide electrolysis cell or SOEC system into a reversible system that operates at 30 kW in electrolysis mode and approximately 10 kW in fuel cell mode. The reversible SOC system will be operated for over 3,000 hours using an improved catalyst in the fuel electrode. Steam for the electrolysis will come from an electric boiler Thermal Energy Distribution System that will be configured to mimic an industrial source of low-grade heat. Thermodynamic analysis will be performed to demonstrate the potential of the system to achieve >85% system efficiency in electrolysis mode. Finally, a technoeconomic analysis will be completed to show potential to produce hydrogen at a cost of $2/kg. The figure at the right indicates a target cost breakdown to achieve that goal.

08 HYDROGEN↗
DOE OSTI2024

Elucidating the corrosion mechanism of commercial Ni-based Superalloys in UCl3 containing-chloride Molten Salt Systems

Molten salt reactors (MSRs) have gained renewed interest, providing several advantages over their predecessors, including the capability to consume spent fuels, enhancing the environmental sustainability of the uranium fuel cycle. For example, molten chloride fast reactors (MCFRs) can reach criticality with molten chloride spent fuel containing high concentrations of impurities, such as actinide products like uranium chloride (UCl3). However, the redox potential of chloride molten salt fuels may change in the presence of these impurities, dictating their corrosivity and in turn the corrosion performance of structural components, such as those constructed from nickel (Ni)-based alloys. The purpose of this investigation is to assess the extent of corrosion of Ni-based alloy, Inconel 617, when exposed to UCl3-LiCl-KCl eutectic salt. Inconel 617 one of only six structural materials that are fully qualified by the American Society for Mechanical Engineers (ASME) Boiler and Pressure Vessel Code for high-temperature nuclear reactor components, making it a technologically mature material to consider for constructing MCFRs. Inconel 617 specimens were submerged in a static LiCl-KCl-UCl3 eutectic salt mixture heated at 700 C for 1000 h under an inert atmosphere. Upon completion, the extent of corrosion was analyzed through a multi-modal characterization approach spanning the engineering to nanoscale, employing computed tomography, focused-ion beam, and transmission electron microscopy techniques. Results from this investigation will enhance our understanding of property-to-performance relationships of candidate structural materials for MSRs with respect to corrosion resistance and interactions between the salt and alloy interface.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Segregation of Chromium and Titanium in Sapphire Optical Fiber Grown via the Laser-Heated Pedestal Growth Technique

Our research involves growth of single crystal (SC) optical fibers to be used for sensing applications in harsh environments. Silica optical fibers are an affordable and reliable option for a wide variety of applications including optical fiber sensors and fiber lasers. However, for applications in harsh environments, such as high temperatures, radioactivity, corrosivity, etc., silica fibers are not suitable due to their instability under such conditions. Fibers composed of SC materials such as sapphire and YAG are mechanically, chemically, and thermally more robust to harsh conditions, and thus are more appropriate for sensing applications in environments such as nuclear reactors, jet engines, and boiler. However, SC fibers grown via the laser-heated pedestal growth (LHPG) technique do not intrinsically have a functional cladding layer. A cladding layer is required to reduce the modal volume for distributed sensing applications, to reduce frustrated total internal reflection induced by surface contact of the fiber in certain applications, and to improve transmissivity. Our lab investigates introduction of dopant materials during LHPG to induce an effective core-cladding structure while maintain the crystallinity of the host material. This process results in optical fiber that is not only robust to harsh environments, but also has improved optical properties for distributed sensor applications.

distributed sensing↗
DOE OSTI2025

RIM: Reliability and Integrity Management/ASME BPVC Section XI Division 2

Prepared at the request of the Nuclear Energy Institute (NEI) and the Joint Committee on Nuclear Risk Management's SubCommittee on Risk Applications, this slide set presents an update on some recent Reliability and Integrity Management (RIM) Subgroup work to respond to industry questions on American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI Division 2. If approved, the briefing will be given at an upcoming workshop of the Nuclear Energy Institute (February 18-20, Washington, DC) and at an upcoming meeting of the Joint Committee on Nuclear Risk Management's SubCommittee on Risk Applications (February 25-26, College Park, MD).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Energy Saving Estimation of ASHRAE Guideline 36 Supervisory Setpoint Reset Controls in a Commercial Large Office Building

Designing, commissioning, and retrofitting HVAC control systems for energy efficiency is crucial, but the use of ad-hoc control sequences by designers and contractors, based on scattered information, results in diverse and sub-optimal sequences. ASHRAE Guideline 36 (G36) addresses the challenge by providing standardized, rule-based HVAC control sequences that prioritize energy efficiency. However, there is limited evaluation of their energy performance at the building level, with only a few studies primarily focused on HVAC airside systems in small-to-medium-sized commercial buildings. In this study, the energy performance of ASHRAE Guideline 36 control sequences was assessed using a large office building emulator in Chicago. The emulator features a central plant system with multiple chillers and boilers as well as multiple variable air volume (VAV) systems with terminal reheat. To achieve a high-fidelity representation, we developed a Spawn-of-EnergyPlus-based model for the large office building, maintaining the DOE prototype large office building setup but substituting the HVAC system with its Modelica counterpart. This substitution ensures that the building thermal load, HVAC system's dynamics, and detailed control sequences are all accurately represented. The study involved evaluating and implementing control strategies outlined in ASHRAE Guideline 36-2021 to replace conventional controls. These strategies include the demand-based supply air temperature and duct static pressure setpoint reset and the request logic for demand-based reset of chilled/hot water supply temperature setpoints and pipe static pressure setpoints. Energy performance was evaluated under various load conditions, including cooling, heating, and transitional seasons, both for individual control strategies and in combination. The results indicate that the collective control strategies retrofit yield greater energy savings than the sum of individual strategies, highlighting the synergistic benefits of incorporating both airside and plant-side control retrofits. Additionally, energy savings of up to 41% in the heating season, 18% in the shoulder season, and 20 % in the cooling season were observed compared to baseline control while maintaining the thermal comfort level.

ASHRAE Guideline 36, Commercial buildings, Control↗
DOE OSTI2024

Fitness-for-Service Analysis of Reactor Components under Flexible Load-Following Operating Conditions

Conventional power-generation plants, including nuclear plants, have been traditionally designed to provide a steady baseload energy capacity, optimizing output efficiency while minimizing variable costs. However, the growing adoption of large-scale renewable energy-generation systems, which rely on intermittent sources such as solar and wind, has introduced more variability into the energy supply in interconnected electricity grids. As a result, the next generation of power plants needs to operate in what is known as the load-following mode, requiring flexible adjustments in electricity production to align with the energy demand on the grid. This transition from the steady baseload operation to load-following operating conditions can significantly increase the number of times various plant components are exposed to transient stresses. This increased thermo-mechanical cycling can lead to accelerated material degradation, thereby elevating the risk of premature failure of a component. It becomes imperative to conduct a comprehensive analysis of fatigue, creep-fatigue, and stress corrosion cracking life, to assess the resilience of the various engineering components under these flexible load-following operating conditions. This study aims to develop a comprehensive numerical model of a light-water reactor pressure vessel (RPV) to investigate its degradation under various operating scenarios. This coupled thermo-mechanical finite element analysis evaluated the stress response of the RPV caused by considering fluctuations in thermal and mechanical loads caused by the varying pressure and temperature occurring during the load-following operation. Critical locations on the RPV are subsequently identified based on the stress response. The stress intensity factors for the postulated flaws at those locations are then calculated, followed by an evaluation of the reactor's life in accordance with the ASME Boiler and Pressure Vessel Code Section XI. This comprehensive life assessment covers a number of transients expected during the flexible load-following operation, providing invaluable insights into the RPV's structural integrity. Moreover, the development methodology can be adapted to other reactor components, as well as components of conventional power stations that are affected by varying operating conditions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Techno-Economic Analysis of Sorbent-Based Direct Air Capture Informed by EPC Input and Recent Technological Advancements

Direct air capture (DAC) technology has been developing at an accelerated rate. Independent, comprehensive, and transparent techno-economic analyses (TEAs) of DAC technologies are required to inform future development. NETL is publishing an update to the 2022 sorbent-based DAC case study report. This manuscript highlights select TEA results that will be presented in the upcoming NETL report. The analysis considers three sorbent-based DAC system configurations (direct steam temperature swing, indirect steam temperature swing, and direct steam temperature/vacuum swing) and two scenarios for providing the required electricity and steam (stand-alone system with a combined heat and power plant generating the required steam and electricity, and a purchased power case utilizing an electric boiler). Detailed sensitivity analyses are included and highlight the impact of external market factors, sorbent performance, and economies of scale. The potential for cost reductions by opting for temperature vacuum swing adsorption is highlighted. More detailed background and TEA results will be presented in the upcoming NETL report titled “Direct Air Capture Case Studies: Sorbent System (Revision 1)”.

carbon dioxide removal↗
DOE OSTI2025

An ASME-Compliant Helium-4 Evaporation Refrigerator for the SpinQuest Experiment

This paper presents the design, safety basis, and commissioning results of a 1 K liquid helium-4 (4He) evaporation refrigerator developed for the Fermilab SpinQuest Experiment (E1039). The system represents the first high power helium evaporation refrigerator operated in a fixed target scattering experiment at Fermilab and was engineered to comply with the Fermilab ES&H Manual (FESHM) requirements governing pressure vessels, piping, cryogenic systems, and vacuum vessels. The design is mapped to ASME B31.3 (Process Piping) and the ASME Boiler and Pressure Vessel Code (BPVC) for pressure boundary integrity and overpressure protection, with documented compliance to FESHM Chapters 5031 (Pressure Vessels), 5031.1 (Piping Systems), and 5033 (Vacuum Vessels). This work documents the methodology used to reach compliance and approval for the 4He evaporation refrigerator at Fermilab which the field lacks. Design considerations specific to the high radiation target-cave environment including remotely located instrumentation approximately 20 m from the cryostat are summarized, together with the relief-system sizing methodology used to accommodate transient heat loads from dynamic nuclear polarization microwaves and the high-intensity proton beam. Commissioning data from July 2024 confirms that the system satisfies all thermal performance and safety objectives.

Roberts, Jordan D. [Virginia U.]↗
DOE OSTI2024

Evaluation of the Simplified Assessment Peak Equivalent Stress Design Limit Probability of Failure

The ASME Boiler and Pressure Vessel Code Section III Division 5 Subsection HH Subpart A (HHA) outlines requirements and qualification methods for graphite core components in high temperature reactor environments. One such structural qualification method is the Simplified Assessment outlined in HHA-3220, which represents a comparison of service level specific equivalent stresses to allowable stress values derived from target probability of failures (POFs). The target POF is based on the structural reliability class (SRC) for the components and service condition of interest, and is used to establish design limits for combined membrane and peak equivalent stresses. Combined membrane stress limits are derived from two parameter Weibull distribution fits of the graphite grade specific tensile strength data and the target POF. The combined membrane stress limits are then scaled based on the ratio of flexural strength to tensile strength (referred hereinafter as Rtf) to establish the corresponding peak equivalent stress limits. It is desired that the target POF is at least maintained for both combined membrane and peak equivalent design limits, especially with the use of Rtf. The assessment presented herein examines tensile and flexural data for medium and fine grade graphites relative to the calculated combined membrane and peak equivalent stress limits per HHA-3220 and illustrates the simplified assessment method results in anticipated, or at least conservative, peak equivalent design stress limits.

36 - MATERIALS SCIENCE↗
DOE OSTI2025

NiCrMoNb age hardenable alloy for creep-resistant high temperature applications, and methods of making

Nickel alloys, methods of making nickel alloys, articles including the nickel alloys, uses of the alloys, and methods of treating nickel alloys are described. The inventive heat resistant structural materials are suitable for applications requiring high yield stress at room temperature and good creep strength at high temperatures, such as in gas turbines, steam turbines, fossil energy boilers, aero engines, power generation systems using fluids such as supercritical carbon dioxide (e.g., advanced ultra-supercritical power plants), concentrated solar power plants, nuclear power plants, molten salt reactors: turbine blades, casings, valves, heat exchangers and recuperators.

Detrois, Martin↗