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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.

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At least 379 records · Page 21
All sourcesNASA NTRSDOE OSTI
NASA NTRS1991

A dual-cooled hydrogen-oxygen rocket engine heat transfer analysis

The potential benefits of simultaneously using hydrogen and oxygen as rocket engine coolants are described. A plug-and-spool rocket engine was examined at heat fluxes ranging from 9290 to 163,500 kW/sq m, using a combined 3-D conduction/advection analysis. Both counter flow and parallel flow cooling arrangements were analyzed. The results indicate that a significant amount of heat transfer to the oxygen occurs, reducing both the hot side wall temperature of the rocket engine and also reducing the exit temperature of the hydrogen coolant. In all heat flux and coolant flow rates examined, the total amount of heat transferred to the oxygen was found to be largely independent of the oxygen coolant flow direction. At low heat flux/low coolant flow (throttled) conditions, the oxygen coolant absorbed more than 30 percent of the overall heat transfer from the rocket engine exhaust gasses. Also, hot side wall temperatures were judged to decrease by approximately 120 K in the throat area and up to a 170 K combustion chamber wall temperature reduction is expected if dual cooling is applied. The reduction in combustion chamber wall temperatures at throttled conditions is especially desirable since tha analysis indicates that a double temperature maxima, one at the throat and another in the combustion chamber, occurs with a traditional hydrogen cooled only engine. Conversely, a dual cooled engine essentially eliminates any concern for overheating in the combustion chamber.

Kacynski, Kenneth J.↗
NASA NTRS1995

Hydrogen film cooling with incident and swept-shock interactions in a Mach 6.4 nitrogen free stream

The effectiveness of slot film cooling of a flat plate in a Mach 6.4 flow with and without incident and swept oblique shock interactions was experimentally investigated. Hydrogen was the primary coolant gas, although some tests were conducted using helium as the coolant. Tests were conducted in the Calspan 48-Inch Shock Tunnel with a nitrogen flow field to preclude combustion of the hydrogen coolant gas. A two-dimensional highly instrumented model developed in a previous test series was used. Parameters investigated included coolant mass flow rate, coolant gas, local free-stream Reynolds number, incident oblique shock strength, and a swept oblique shock. Both gases were highly effective coolants in undisturbed flow; however, both incident and swept shocks degraded that effectiveness.

Olsen, George C.↗
NASA NTRS2005

Technical Consultation of the International Space Station (ISS) Internal Active Thermal Control System (IATCS) Cooling Water Chemistry

The Internal Active Thermal Control System (IATCS) coolant exhibited unexpected chemical changes during the first year of on-orbit operation following the launch and activation in February 2001. The coolant pH dropped from 9.3 to below the minimum specification limit of 9.0, and re-equilibrated between 8.3 and 8.5. This drop in coolant pH was shown to be the result of permeation of CO2 from the cabin into the coolant via Teflon flexible hoses which created carbonic acid in the fluid. This unexpected diffusion was the result of having a cabin CO2 partial pressure higher than the ground partial pressure (average 4.0 mmHg vs. less than 0.2 mmHg). This drop in pH was followed by a concurrent increasing coolant nickel concentration. No other metal ions were observed in the coolant and based on previous tests, the source of nickel ion was thought to be the boron nickel (BNi) braze intermetallics used in the construction of HXs and cold plates. Specifically, BNi2 braze alloy was used for the IATCS IFHX and BNi3 braze alloy was used for the IATCS Airlock Servicing and Performance Checkout Unit (SPCU) HX and cold plates. Given the failure criticality of the HXs, a Corrosion Team was established by the IATCS CWG to determine the impact of the nickel corrosion on hardware performance life.

Gentz, Steven J.↗
NASA NTRS2006

Microbial Characterization of Internal Active Thermal Control System (IATCS) Hardware Surfaces after Five Years of Operation in the International Space Station

A flex hose assembly containing aqueous coolant from the International Space Station (ISS) Internal Active Thermal Control System (IATCS) consisting of a 2 foot section of Teflon hose and quick disconnects (QDs) and a Special Performance Checkout Unit (SPCU) heat exchanger containing separate channels of IATCS coolant and iodinated water used to cool spacesuits and Extravehicular Mobility Units (EMUS) were returned for destructive analyses on Shuttle return to flight mission STS-114. The original aqueous IATCS coolant used in Node 1, the Laboratory Module, and the Airlock consisted of water, borate (pH buffer), phosphate (corrosion control), and silver sulfate (microbiological control) at a pH of 9.5 +/- 0.5. Chemical changes occurred after on-orbit implementation including a decrease to pH 8.4 due to the diffusion of carbon dioxide through the Teflon hoses, an increase in nickel ions due to general corrosion of heat exchanger braze coatings, a decrease in phosphate concentration due to precipitation of nickel phosphate, and the rapid disappearance of silver ions due to deposition on hardware surfaces. Also associated with the coolant chemistry changes was an increase in planktonic microorganisms from less than 100 colony forming units (CFU) per 100 ml to approximately 1 million CFU per 100 ml. Attachment and growth of microorganisms to the system surfaces (biofilm) was suspected due to the levels of planktonic microorganisms in the coolant. Biofilms can reduce coolant flow, reduce heat transfer, amplify degradation of system materials initiated by chemical corrosion, and enhance mineral scale formation.

Roman, Monsi C.↗
NASA NTRS2009

Transient Modeling and Analysis of a Metabolic Heat-Regenerated Temperature Swing Adsorption (MTSA) System for a PLSS

A Metabolic heat-regenerated Temperature Swing Adsorption (MTSA) system is being developed for carbon dioxide, water and thermal control in a lunar and martian portable life support system (PLSS). A previous system analysis was performed to evaluate the impact of MTSA on PLSS design. That effort was Mars specific and assumed liquid carbon dioxide (LCO2) coolant made from martian resources. Transient effects were not considered but rather average conditions were used throughout the analysis. This effort takes into further consideration the transient effects inherent in the cycling MTSA system as well as assesses the use of water as coolant. Standard heat transfer, thermodynamic, and heat exchanger methods are presented to conduct the analysis. Assumptions and model verification are discussed. The tool was used to perform various system studies. Coolant selection was explored and takes into account different operational scenarios as the minimum bed temperature is driven by the sublimation temperature of the coolant (water being significantly higher than LCO2). From this, coolant mass is sized coupled with sorbent bed mass because MTSA adsorption performance decreases with increasing sublimation temperature. Reduction in heat exchanger performance and even removal of certain heat exchangers, like a recuperative one between the two sorbent beds, is also investigated. Finally, the coolant flow rate is varied over the cycle to determine if there is a more optimal means of cooling the bed from a mass perspective. Results of these studies and subsequent recommendations for system design are presented.

Iacomini, Christie↗
NASA NTRS2011

Evaluation of a Passive Heat Exchanger Based Cooling System for Fuel Cell Applications

Fuel cell cooling is conventionally performed with an actively controlled, dedicated coolant loop that exchanges heat with a separate external cooling loop. To simplify this system the concept of directly cooling a fuel cell utilizing a coolant loop with a regenerative heat exchanger to preheat the coolant entering the fuel cell with the coolant exiting the fuel cell was analyzed. The preheating is necessary to minimize the temperature difference across the fuel cell stack. This type of coolant system would minimize the controls needed on the coolant loop and provide a mostly passive means of cooling the fuel cell. The results indicate that an operating temperature of near or greater than 70 C is achievable with a heat exchanger effectiveness of around 90 percent. Of the heat exchanger types evaluated with the same type of fluid on the hot and cold side, a counter flow type heat exchanger would be required which has the possibility of achieving the required effectiveness. The number of heat transfer units required by the heat exchanger would be around 9 or greater. Although the analysis indicates the concept is feasible, the heat exchanger design would need to be developed and optimized for a specific fuel cell operation in order to achieve the high effectiveness value required.

Colozza, Anthony J.↗
NASA NTRS2017

Close Range Photogrammetry in Space - Measuring the On-Orbit Clearance between Hardware on the International Space Station

When photogrammetrists read an article entitled "Photogrammetry in Space" they immediately think of terrestrial mapping using satellite imagery. However in the last 19 years the roll of close range photogrammetry in support of the manned space flight program has grown exponentially. Management and engineers have repeatedly entrusted the safety of the vehicles and their crews to the results of photogrammetric analysis. In February 2010, the Node 3 module was attached to the port side Common Berthing Mechanism (CBM) of the International Space Station (ISS). Since this was not the location at which the module was originally designed to be located on the ISS, coolant lines containing liquid ammonia, were installed externally from the US Lab to Node 3 during a spacewalk. During mission preparation I had developed a plan and a set of procedures to have the astronauts acquire stereo imagery of these coolant lines at the conclusion of the spacewalk to enable us to map their as-installed location relative to the rest of the space station. Unfortunately, the actual installation of the coolant lines took longer than expected and in an effort to wrap up the spacewalk on time, the mission director made a real-time call to drop the photography. My efforts to reschedule the photography on a later spacewalk never materialized, so rather than having an as-installed model for the location of coolant lines, the master ISS CAD database continued to display an as-designed model of the coolant lines. Fast forward to the summer of 2015, the ISS program planned to berth a Japanese cargo module to the nadir Common Berthing Mechanism (CBM), immediately adjacent to the Node 3 module. A CAD based clearance analysis revealed a negative four inch clearance between the ammonia lines and a thruster nozzle on the port side of the cargo vehicle. Recognizing that the model of the ammonia line used in the clearance analysis was "as-designed" rather than "as-installed", I was asked to determine the real clearance between the ammonia lines and expected position of the thruster bell using existing on-orbit imagery. Imagery of the area of interest, taken several years earlier from the Space Shuttle during a fly-around of the ISS, was found and used to set a stereo pair. Space Vision System Targets and Handrail bolts measured in the ISS analytical coordinate system (ISSACS) prior to launch, were used to obtain an absolute orientation so all photogrammetric measurement's would be in the ISSACS coordinate system. Coordinates for the design location of the edges of the thruster bell, when the cargo vehicle was fully berthed to the ISS, were displayed in 3-D relative to the as-installed ammonia lines. This immediately revealed a positive clearance, which was later quantified to be a minimum of 10" +/0.5". The analysis was completed over a single weekend by a single analyst. Using updated imagery, acquired from the station's robotic arm, a complete as-installed model of the coolant lines was generated from stereo photography and replaced the design model in the master ISS CAD database.

Liddle, Donn↗
DOE OSTI2023

Cooling assembly for a turbine assembly

A cooling assembly includes a coolant chamber disposed inside an airfoil of a turbine assembly that directs coolant inside the airfoil. The airfoil extends between a leading edge and a trailing edge along an axial length of the airfoil. Inlet cooling channels are fluidly coupled with the coolant chamber and direct the coolant in a direction toward a trailing edge chamber of the airfoil. The trailing edge chamber is fluidly coupled with at least one inlet cooling channel. The trailing edge chamber is disposed at the trailing edge of the airfoil and includes an inner surface. The inlet cooling channels direct at least a portion of the coolant in a direction toward the inner surface of the trailing edge chamber. One or more outlet cooling channels direct at least a portion of the coolant in one or more directions away from the trailing edge chamber.

Rathay, Nicholas William↗
DOE OSTI2024

Waste heat recovery power electronics cooling

A cooling system for an electrified vehicle includes a first cooling loop for circulating coolant for cooling at least one of power electronics and a motor/generator of the vehicle. The first coolant loop includes a heat exchanger for exchanging heat with the coolant in the first cooling loop. A second cooling loop is provided for circulating coolant for cooling a battery of the vehicle. The second cooling loop includes a coolant chiller connected to a refrigeration system of the vehicle for exchanging heat in the coolant received from the battery with the refrigeration system of the vehicle.

Ernst, Timothy C.↗
DOE OSTI2023

Fluid structural interaction simulation of the MP-1 plate performance irradiated in the advanced test reactor

Due to the high neutron fluxes they generate and increased heat transfer performance, plate type fuels are used in the U.S. high-performance research reactors. During irradiation, a significant amount of fission energy (i.e., ~200 MeV per atom) is released by the U-235 chain reaction that is carried away by the coolant. Through thermal hydraulic analysis, the coolant’s heat transfer capability is investigated to ensure that the desired fuel temperature can be maintained. At high temperatures, the plate undergoes elastic/plastic deformation, creep, and swelling as a result of both the temperature gradients and the fission gas production within the fuel. These effects are studied via fuel performance analysis. When the plate deformation is small enough that the coolant flow remains relatively unchanged, conducting these two types of analyses independently will suffice. But at high fuel burnups, the swollen plates may encroach into the coolant channels that separate the fuel plates from each other and cause these channels to narrow. Large reductions in channel gap size imperil cooling performance, causing fuel temperatures to rise. Further, if the plate deformation is asymmetric, the fuel centerline will shift toward one side of the channel, causing an uneven reduction in coolability. In addition, a boehmite (oxide) layer will, over time, grow on the plate surface, further obstructing the heat transfer process. To precisely predict fuel temperatures/deformation, a complete coupled analysis that considers coolant flow, heat transfer, oxide growth, elastic/plastic deformation, creep, and swelling is needed; however, this type of analysis method is not available in the literature. To fill the gap, this research developed a fluid structure interaction (FSI) approach to the fuel plate analysis, then successfully applied it to the Mini-Plate (MP)-1 experiment, which was irradiated in the Advanced Test Reactor (ATR) for both one and two cycles. The complete analysis coupled STAR-CCM+, a computational fluid dynamic (CFD) software for calculating flow, with Abaqus, a finite element analysis code for calculating plate deformation. Improvements in the results were found when comparing the fully coupled analysis to the independently conducted analyses but they were not significant due to the miniature size of the plates and the relatively short irradiation time. In the future, the fully coupled approach presented herein will be applied to full-size fuel plates with longer irradiation cycles once additional experiments become available.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Code Benchmark of the HTTF Pressurized Conduction Cooldown Test Using SAM

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

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

A User Guide to PARET/ANL

PARET was originally created in 1969 at what is now Idaho National Laboratory (INL), to analyze reactivity insertion events in research and test reactor cores cooled by light or heavy water, with fuel composed of either plates or pins. The use of PARET is also appropriate for fuel assemblies with curved fuel plates when their radii of curvatures are large with respect to the fuel plate thickness. The PARET/ANL version of the code has been developed at Argonne National Laboratory (ANL) under the sponsorship of the U.S. Department of Energy/NNSA since the inception of the Reactor Conversion Program. Since Reduced Enrichment for Research and Test Reactors (RERTR) began in 1978, PARET/ANL has been benchmarked to experimental data including SPERT testing, and used to determine the expected transient behavior of a large number of reactors regardless of enrichment both inside and outside the Reactor Conversion Program. This document provides the pertinent information for the use of PARET/ANL Version 7.6. PARET/ANL models the various fueled regions of a reactor core as channels. Each of these channels consists of a single flat fuel plate/pin (including cladding and, optionally, a gap) with water coolant on each side. In slab geometry the coolant channels for a given fuel plate are of identical dimensions (mirror symmetry), but they can be of different thickness in each channel. There can be many channels, but each channel is independent and coupled only through reactivity feedback effects to the whole core. The time-dependent differential equations that represent the system are replaced by an equivalent set of finite-difference equations in space and time, which are integrated numerically. PARET/ANL uses fundamentally the same numerical scheme as RELAP5 for the time-integration of the point-kinetics equations. The one-dimensional thermal-hydraulic model includes temperature-dependent thermal properties of the solid materials, such as heat capacity and thermal conductivity, as well as the transient heat production and heat transfer from the fuel meat to the coolant. Temperature- and pressure-dependent thermal properties of the coolant such as enthalpy, density, thermal conductivity, and viscosity are also used in determining parameters such as friction factors and heat transfer coefficients. The code first determines the steady-state solution for the initial state. Then the solution of the transient is obtained by integration in time and space. Multiple heat transfer, DNB and flow instability correlations are available. The code was originally developed to model reactors cooled by an open loop, which was adequate for rapid transients in pool-type cores. An external loop model appropriate for Miniature Neutron Source Reactors (MNSR’s) was also added to PARET/ANL to model natural circulation within the vessel, heat transfer from the vessel to pool and heat loss by evaporation from the pool. PARET/ANL also contains models for decay heat after shutdown, control rod reactivity versus time or position, time-dependent pump flow, and loss-of-flow event with flow reversal as well as logic for trips on period, power, and flow. Feedback reactivity effects from coolant density changes and temperature changes are represented by tables. Feedback reactivity from fuel heat-up (Doppler Effect) is represented by a four-term polynomial in powers of fuel temperature. Photo-neutrons produced in beryllium or in heavy water may be included in the point-kinetics equations by using additional delayed neutron groups.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Coupling Thermal-Hydraulics with Reactor Pressure Vessel Fracture Models

Because of the importance of maintaining the integrity of the reactor pressure vessel (RPV) in light-water reactors, significant effort has been devoted over the last several decades to understand the potential failure mechanisms, characterize how material properties evolve due to environmental conditions, and develop tools to assess the probability of fracture in aged RPVs during potential transient events. One important aspect of this problem that has not yet been fully addressed is the effects of nonuniform cooling on the RPV inner wall due to the way low-temperature water injected into the primary coolant system is distributed within the RPV during potential loss-of-coolant accident transients. Colder temperatures are expected in the regions near the inlets, causing a “cold-plume” effect. Rapid cooling of the RPV wall increases the likelihood of fracture initiation, and localized cooling due to cold-plume effects could potentially exacerbate this problem. The Grizzly code can perform a three-dimensional simulation of the RPV thermomechanical response as well as the probabilistic assessment of fracture for a population of pre-existing flaws, which is necessary to account for spatial variations in coolant temperature. This report documents a first effort to couple Grizzly RPV fracture models with computational fluid dynamics models performed using the Cardinal code, which allows for a realistic representation of the temperature distribution. This report demonstrates this coupling approach on two test cases: a simplified proof-of-concept scenario and a realistic small-break loss-of-coolant accident scenario. In both cases, the effects of the spatially varying temperature are evident; although, because the small-break loss-of-coolant accident scenario was far from challenging the RPV’s margins, it is still inconclusive how important cold-plume effects are in realistic accident scenarios. However, this effort represents an important step toward performing realistic analyses to determine to what extent spatially nonuniform cooling might affect RPV integrity under more aggressive accident scenarios.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Integrated Turbine Component Cooling Designs Facilitated by Additive Manufacturing and Optimization

This research program involved the development of improved gas turbine cooling by optimizing the combined internal and film cooling configurations used for turbine blades and vanes. Improved cooling of turbine airfoils allows the gas turbines to be operated at higher temperatures which improves efficiency for gas turbine-based power generation. This also facilitates operation with hydrogen-based fuels which inherently generate higher operating temperatures. Designs of the new cooling configurations were based on the added geometric flexibility that is available when using metal additive manufacturing techniques. When making these designs, adjustments were made to account for imperfections occurring in the metal additive manufacturing process. New designs for film cooling holes were developed using computational adjoint gradient optimization techniques which generated a unique film cooling hole with 70% greater film cooling effectiveness compared to conventional shaped film cooling holes. A major factor in improving the film cooling effectiveness were small external protrusions on either side of the new hole which generated vortices that pushed coolant towards the surface and increased the lateral spreading of the coolant. Our studies also included optimization of film cooling holes that can be built with conventional manufacturing techniques and optimizing the internal coolant channels that supply coolant to the film cooling holes. Various internal coolant channel shapes were investigated and designs that maximized the cooling of the airfoil while minimizing pressure losses in the channel flows were identified. Testing and verification of the performance of the final configurations was done with a transonic wind tunnel facility which allowed testing at engine realistic high velocities. Final testing of the combined internal and film cooling configurations was done by incorporating the configurations in Penn State’s National Experimental Turbine (NExT) vane. These tests confirmed the enhanced overall cooling effectiveness provided by the new cooling configuration designs.

03 NATURAL GAS↗
DOE OSTI2025

The SAS4A/SASSYS-1 Version 5.8 Safety Analysis Code System

SAS4A/SASSYS-1 is a software simulation tool used to perform deterministic analysis of anticipated events as well as design basis and beyond design basis accidents for advanced nuclear reactors. Detailed, mechanistic models of steady-state and transient thermal, hydraulic, kinetic, and mechanical phenomena are employed to describe the response of the reactor core, the reactor primary and secondary coolant loops, the reactor control and protection systems, and the balance-of-plant to accidents caused by changes in coolant flow, loss of heat rejection, or reactivity insertion. The consequences of single and double-fault accidents can be modeled, including fuel and coolant heating, fuel and cladding mechanical behavior, core reactivity feedbacks, coolant loop performance including natural circulation, and decay heat removal. Analyses are typically terminated upon demonstration of reactor and plant shutdown to permanently coolable conditions, or upon violation of design basis margins. The objective of the analysis is to quantify accident consequences as measured by the transient behavior of system performance parameters, such as fuel and cladding temperatures, reactivity, and cladding strain. Originally developed for analysis of sodium cooled reactors with oxide fuel clad by stainless steel, the models were subsequently extended and specialized to metallic fuel clad with advanced alloys and to several other coolant options, including lead, LBE, and water.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Gas-Gap Calorimeter Sizing and Rating Tools for Heat Pipe Experimentation

Effective gas-gap calorimeter sizing and rating tools are required to design calorimeters that enable well-defined operating conditions, high cooling powers, and accurate calorimetry. The present report describes two tools utilizing Microsoft Excel and MathWorks MATLAB that enable the sizing and rating of gas-gap calorimeters with a He-Ar binary gas mixture in the gap. The Excel tools are two easy-to-use spreadsheets for calculating heat pipe operating temperature or He-Ar mole fractions given the required inputs. The MATLAB tool includes similar models with more robust solution algorithms along with sub-options for full-vacuum, static gas-gap, and forced circulation in the gas-gap. The accuracy of the developed tools were demonstrated by comparing with existing work in literature. The MATLAB tool was used for a parametric study to determine the gas-gap thickness, coolant gap thickness, coolant flow rate, and coolant inlet temperatures for the testing of a 3/4 in outer diameter heat pipe up to powers of 10 kW and temperatures of 1,000°C. The effects of heat pipe and calorimeter inner shell surface emissivities were investigated, considering the inability to control or accurately measure surface emissivities in most applications. It was found that controlling the He-Ar mole fractions provides a wide operating range for gas-gap thicknesses around ~ 0.042--0.090 in (~ 1.07--2.29 mm) for a surface emissivity range of 0.4--0.8, since conduction heat transfer is a significant fraction of the heat transfer across the gas-gap. In addition, it was found that turbulent flow in the coolant gap is needed at high input powers to prevent shell temperatures from approaching the boiling temperature of the water coolant. The parametric study resulted in choosing stainless steel tubes with an outer diameter and thickness of 1 x 0.065 in as the inner shell, and a 1-1/2 x 0.156 in as the outer shell of the calorimeter. Overall, this report presents the necessary information for the sizing and rating of gas-gap calorimeters with He-Ar mixtures for the testing of high-temperature heat pipes (~ 500--1,000°C).

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Reassessing Double-Ended Guillotine Break Requirements: Evidence-Based Analysis of Regulatory Assumptions After Five Decades of Nuclear Operation

After five decades of nuclear power operation encompassing more than 20,000 reactor-years across 35 countries and 647 reactors, zero double-ended guillotine breaks (DEGBs) have been documented in commercial reactor coolant systems—despite DEGB being the fundamental design-basis assumption driving Emergency Core Cooling System (ECCS) sizing, structural protection requirements, and containment design specifications. This report examines the basis for DEGB requirements in nuclear power plant design. The DEGB postulate assumes the instantaneous, complete circumferential severance of the largest diameter pipes in reactor coolant systems, driving major design requirements under 10 Code of Federal Regulations 50.46, General Design Criterion 4 and containment design specifications. The United States (4,880 reactor-years) and France (2,505 reactor-years) contribute the largest operational datasets. Probabilistic assessments estimate direct DEGB occurrence probabilities with extremely low event frequencies, far below the 10-5/reactor-year thresholds typically used to define non-credible events in nuclear-safety analyses; i.e., events with probability this low fall into beyond-design-basis events. Current material-science knowledge demonstrates that the ductile steel materials used in nuclear piping systems exhibit stable crack-growth behavior fundamentally incompatible with instantaneous severance. International regulatory experience, particularly Germany’s comprehensive break-preclusion implementation, and successful leak-before-break (LBB) applications in almost all of U.S. pressurized water reactor units validate that alternatives can maintain safety performance while reducing economic burden. Current DEGB protection systems impose estimated lifetime costs of hundreds of millions of dollars per unit, over the life of a plant across the nuclear industry (including ongoing costs), representing substantial resource allocation toward scenarios with extremely low probability. Although this report acknowledges uncertainties regarding long-term aging effects, potential synergistic degradation mechanisms, and site-specific seismic considerations that warrant continued evaluation as regulatory policy evolves, there remains no documented evidence that a DEGB has occurred as a consequence of the conditions or mechanisms described in this report. This report acknowledges the Nuclear Regulatory Commission’s (NRC’s) recent efforts—outlined in the draft Interim Staff Guidance (ISG) NRC-DSS-ISG-2025-XX (“Treatment of Certain Loss-of-Coolant Accident Locations as Beyond-Design-Basis Accidents Draft Interim Staff Guidance”)—to reduce overly conservative requirements for large-break loss of coolant accidents through technical justifications and exemptions. However, extensive operating experience and validated methodologies—such as LBB and in-service inspection programs—demonstrate that the probability of a DEGB in reactor coolant-loop piping is extremely low, even under seismic conditions. The authors and reviewers of this report recommend that DEGB be removed as a design-basis event through formal rulemaking, rather than case-by-case exemptions, to better reflect credible failure modes, align with current data, and align with modern, risk-informed safety analysis.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1973

Cold-air aerodynamic study in a two-dimensional cascade of a turbine stator blade with suction-surface film cooling

The effect on aerodynamic performance of a single row of spanwise-spaced coolant holes located at four positions chordwise along the suction surface was investigated. In addition, multiple-row data were obtained. The data are presented in terms of primary efficiency, coolant-to-primary-flow percentage, and coolant pressure ratio. Primary efficiencies of multirow blade configurations compare satisfactorily with that calculated from the single-row efficiency increments. At any given coolant flow percentage, the efficiency was about the same for all row locations. For a given coolant pressure, the efficiency varied by as much as 1 percent, depending on the row location.

Brown, D. B.↗