Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “RADIATION PRESSURE”

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

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

At least 307 records · Page 17
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Optimization of microwave emission from laser filamentation with a machine learning algorithm

We demonstrate that is it possible to optimize the yield of microwave radiation from plasmas generated by laser filamentation in atmosphere through manipulation of the laser wavefront. A genetic algorithm controls a deformable mirror that reconfigures the wavefront using the microwave waveform amplitude as feedback. Optimization runs performed as a function of air pressure show that the genetic algorithm can double the microwave field strength relative to when the mirror surface is flat. Here, an increase in the volume and brightness of the plasma fluorescence accompanies the increase in microwave radiation, implying an improvement in the laser beam intensity profile through the filamentation region due to the optimized wavefront.

Englesbe, Alexander↗
DOE OSTI2020

Fully compressible simulations of waves and core convection in main-sequence stars

Context: Recent, nonlinear simulations of wave generation and propagation in full-star models have been carried out in the anelastic approximation using spectral methods. Although it makes long time steps possible, this approach excludes the physics of sound waves completely and requires rather high artificial viscosity and thermal diffusivity for numerical stability. A direct comparison with observations is thus limited. Aims: In this study, we explore the capabilities of our compressible multidimensional Seven-League Hydro (SLH) code to simulate stellar oscillations. Methods: We compare some fundamental properties of internal gravity and pressure waves in 2D SLH simulations to linear wave theory using two test cases: (1) an interval gravity wave packet in the Boussinesq limit and (2) a realistic 3 M ⊙ stellar model with a convective core and a radiative envelope. Oscillation properties of the stellar model are also discussed in the context of observations. Results: Our tests show that specialized low-Mach techniques are necessary when simulating oscillations in stellar interiors. Basic properties of internal gravity and pressure waves in our simulations are in good agreement with linear wave theory. As compared to anelastic simulations of the same stellar model, we can follow internal gravity waves of much lower frequencies. The temporal frequency spectra of velocity and temperature are flat and compatible with the observed spectra of massive stars. Conclusion: The low-Mach compressible approach to hydrodynamical simulations of stellar oscillations is promising. Our simulations are less dissipative and require less luminosity boosting than comparable spectral simulations. The fully-compressible approach allows for the coupling of gravity and pressure waves in the outer convective envelopes of evolved stars to be studied in the future.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2024

Comparison of Americium(III) and Neodymium(III) Monothiophosphate Complexes

Mixed-donor ligands, such as those containing a combination of O/N or O/S, have been studied extensively for the selective extraction of trivalent actinides, especially Am 3+ and Cm3+, from lanthanides during the recycling of used nuclear fuel. Oxygen/sulfur donor ligand combinations also result from the hydrolytic and/or radiolytic degradation of dithiophosphates, such as the Cyanex® class of extractants, that are initially converted to monothiophosphates. To understand potential differences between the binding of such degraded ligands to Nd 3+ and Am 3+ , the monothiophosphate complexes [M(OPS(OEt) 2 ) 5 (H 2 O) 2 ] 2- (M 3+ = Nd 3+ , Am 3+ ) were prepared and characterized by single crystal X-ray diffraction and optical spectroscopy and studied as a function of pressure up ca. 14 GPa using diamond-anvil techniques. Although Nd 3+ and Am 3+ have nearly identical eight-coordinated ionic radii, these structures reveal that while the M-O bond distances in these complexes are nearly equal that the M-S distances are statistically different. Moreover, for [Nd(OPS(OEt) 2 ) 5 (H 2 O) 2 ] 2- , the hypersensitive 4 I 9/2 → 4 G 5 / 2 transition shifts as a function of pressure by -11 cm -1 /GPa. Whereas for [Am(OPS(OEt) 2 ) 5 (H 2 O) 2 ] 2- , the 7 F 0 → 7 F 6 transition shows a slightly stronger pressure dependence with a shift of -13 cm -1 /GPa and also exhibits broadening of the 5f →5f transitions at high pressures. This data likely indicates an increased involvement of the 5f orbitals in bonding to Am 3+ relative to that of Nd 3+ in these complexes.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2025

Experimental test of model predictive control in a variable air volume system

Model predictive control (MPC) has been widely studied as a promising approach for improving energy efficiency and operational flexibility in buildings, yet its real-world performance for commercial variable air volume (VAV) systems remains insufficiently characterized. In particular, the impacts of model mismatch on control robustness, real-time computational burden, and device-level operation are rarely evaluated using long-term field data. Here, this study presents a comprehensive experimental evaluation of MPC applied to a full-scale VAV system in Oak Ridge National Laboratory’s Flexible Research Platform-2 building with constant cooling/heating temperature setpoints and no occupancy. The study offers three key advantages over existing work: (1) it uses a representative building in a full-scale experimental test, capturing realistic system dynamics and complexity; (2) it evaluates a relatively sophisticated MPC formulation using two different optimization solvers (Gurobi and PSO), fully accounting for computational complexity and methodological diversity; and (3) it systematically assesses potential negative impacts on various building devices, benchmark against a well-established baseline, ASHRAE Guideline 36 (G36). To isolate zone- and air-handling-unit–level supervisory control effects, the supply fan was operated with a fixed static pressure setpoint under all strategies, and the trim-and-response static pressure reset in G36 was not enabled. Results show that MPC maintained thermal comfort while improving energy efficiency. Abrupt solar radiation variations degraded performance. Computation times ranged from ∼1 s (Gurobi) to ∼ 70 s (PSO). Compared with G36, MPC achieves 33% energy savings and reduces median reheat coil output by approximately a factor of 5–10 for a representative cooling day under matched weather conditions. However, it increases the maximum discomfort deviation from 0.5 to 1°C and results in a 32% increase in staging frequency. In addition, PSO-based MPC introduced damper oscillations, also affecting actuator longevity.

ASHRAE guideline 36↗
DOE OSTI2020

Protection of graphite from salt and gas permeation in molten salt reactors

The reactor core, moderator and reflectors of a thermal spectrum advanced molten salt reactor will constitute multi-tons of graphite. Porous reactor-grade graphite, if unprotected, can be permeated by molten salt depending on the infiltrating pressure differential and entrance diameters of accessible graphite pores. Salt and gas permeation of graphite can affect microstructural properties and radiation behavior but also facilitate diffusion, deposition and retention of fission products and tritium. Because of the significant void volume of nuclear graphite, fission products and tritium retention due to salt permeation necessitates seal coatings or pore impregnation to reduce open porosity. Alternatively, very fine-grained graphite grades with low Xe permeability are being developed. In this work, we survey the current technologies for mitigating salt and gas transport into graphite.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Precision studies of QCD in the low energy domain of the EIC

This White Paper aims at highlighting the important benefits in the science reach of the EIC. High luminosity operation is generally desirable, as it enables producing and harvesting scientific results in a shorter time period. Furthermore, it becomes crucial for programs that would require many months or even years of operation at lower luminosity.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Benchmarking of ENDF/B-VIII.0 cross sections for LWR ex-core transport applications

ENDF/B-VIII.0 cross section data is benchmarked on a simple Monte-Carlo model representative of the ex-core geometry of a PWR used for reactor vessel neutron fluence calculations. The benchmarking indicates that the ENDF/B-VIII.0 results show large differences as compared to previous evaluations, which have been extensively compared to measured data. This exercise suggests that ENDF/B-VIII.0 nuclear data is not suitable for ex-core LWR radiation transport applications. Evaluators are aware of the problems with the ENDF/B-VIII.0 evaluation of iron and work is underway to improve the evaluations for an upcoming release.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Methodological guidelines on concrete degradation based on predictive models and the release of MOSAIC for industry use

In light-water reactors (LWRs), the concrete biological shield (CBS) designates the concrete structure directly facing the reactor pressure vessel (RPV). The primary function of the CBS is to protect equipment and personnel from the neutron and gamma radiation exiting the RPV. Most CBSs in operation in the United States also provide structural support to the RPV. Based on a long-term structural analysis, the structural integrity of the CBS to transfer the in-service passive load to the foundation, dynamic loading during an earthquake, and thermal loading during a loss-of-coolant accident (LOCA) must be maintained. Based on test reactor data for accelerated conditions, the mechanical properties of irradiated concrete are affected when exposed to neutron fluence higher than ≈ 10 19 n.cm -2 (E > 0.1 MeV). At 80 years of operation, the fluence estimates at the surface of the CBS range between 1×10 19 and 7×10 19 n.cm -2 (E > 0.1 MeV): the fluence value depends mostly on the design. Although fluence attenuates through the CBS by virtue of the shielding properties of concrete, about 5% to 10% of the CBS wall depth is affected by high fluence. The main degradation mechanism is the so-called radiation-induced volumetric expansion (RIVE) caused by the amorphization of aggregate-forming minerals. RIVE causes large deformation of up to 18% in pure quartz, and it also reduces the structural properties of concrete. The structural effects of long-term exposure to fast neutron irradiation in LWRs was studied in the report entitled “Assessment of the Effect of the Irradiation-Induced Degradation on the In-Service Structural Performance of the Concrete Biological Shields” (M3LW-23OR0403013). This report illustrates that (1) irradiation-induced damage expands further than the region subject to a cumulative fluence of 10 19 n.cm -2 (E > 0.1 MeV) to reach the steel reinforcement location, and (2) the irradiation-induced damage region shows no substantial residual bearing capacity and should not be accounted for in the integrity assessment of the CBS.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Radiation Effects in Concrete for Nuclear Systems

Concrete is massively used in the construction of critical safety-related components in nuclear power plants. Among those, the concrete biological shield is exposed to high levels of neutron and gamma irradiation exiting the reactor pressure vessel. This article provides a comprehensive overview of the state-of-the-art knowledge on irradiated concrete. First, the fundamental irradiation-induced degradation mechanisms occurring at the level of the concrete constituents such as rock-forming minerals and hydrated cementitious phases are examined. The critical role of the mineralogical and chemical composition regarding concrete resistance against radiation effects such as amorphization, irradiation-induced expansion and radiolytic gas release are detailed. Second, the interactions between the varied irradiated phases are comprehensively analyzed and put in perspectives with the effects of prolonged irradiation on the structural properties of concrete aggregates, cement paste, and concretes. Finally, the significance of irradiation for the long-term operation of nuclear power plants is discussed in the light of the most recent knowledge and its limitations.

Le Pape, Yann↗
DOE OSTI2025

Conceptual Designs for Irradiation Creep Testing of SiC in HFIR

Understanding irradiation creep of nuclear fuel cladding is important to properly size the initial fuel-cladding gap and understand when pellet-cladding contact is expected to occur due to a combination of fuel swelling and cladding creep-down. Irradiation creep also plays a role in relaxing stresses that develop in-pile. Silicon carbide fiber–reinforced silicon carbide matrix (SiC/SiC) composites are the leading long-term accident-tolerant fuel cladding concept for light-water reactors (LWRs). Although some limited data are available regarding irradiation creep of the individual constituents (fibers, matrix), data regarding irradiation creep of SiC/SiC composites are currently insufficient. Additional data regarding irradiation creep compliance and the rupture lifetime (combination of creep and slow crack growth) are needed to understand material limitations. This work describes the design and development of two irradiation vehicles that are being pursued for testing SiC/SiC concepts in the High Flux Isotope Reactor (HFIR). The first is a passive experiment, referred to as the PRECISE experiment, that leverages the constant coolant pressure of HFIR to compress a metallic bellows and provide a well-characterized load to drive creep in a SiC/SiC dog bone specimen. The total creep strain would be quantified post-irradiation by measuring dimensional changes of the specimen length as well as local dimensional changes within the gauge region. Non-stressed specimens would also be irradiated under the same conditions to provide an indication of dimensional changes due to radiation-induced swelling in the absence of creep. A second, more complex experiment, referred to as the INSITE experiment, is being designed in parallel that would use pneumatics to pressurize a metal bellows and linear variable differential transformers (LVDTs) to measure the specimen displacement in situ during irradiation. Such an experiment would provide significantly more data regarding the evolution of the creep compliance as a function of dose and applied stress within a single experiment but would require significantly more development time and cost to execute. The primary concern with the INSITE experiment is the accuracy, reliability, and expected lifetime of the LVDTs during irradiation at elevated temperatures. Efforts are being made to adjust the experiment design and operating procedure to limit LVDT temperatures and mitigate or otherwise compensate for uncertainties due to factors such as temperature fluctuations, creep in the surrounding structural materials, and drift of the LVDTs. This work describes the experiment designs, thermal and structural analysis that were performed to ensure that the desired temperature and stress conditions can be achieved, some initial sensitivity analyses to predict the evolution of the radiation-induced specimen displacements, and potential sources of uncertainty in the measurements. Out-of-pile testing is being performed in parallel to confirm that the test trains achieve the expected stress states in the specimens and do not result in prohibitive stress concentrators (e.g., in the grip regions) that might risk pre-mature failure. The PRECISE experiments are proceeding toward fabrication and assembly with HFIR insertion planned during fiscal year 2026. The INSITE experiment is progressing toward out-of-pile demonstrations, which will provide more conclusive evidence regarding the feasibility of executing these tests in HFIR or whether alternative displacement monitoring techniques may need to be considered.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Photoelectron spectroscopy of CsK2Sb photocathode at Synchrotron Radiation Facility using vacuum transport system

As accelerators and electron microscopes become more advanced, high-performance photocathodes are required. In particular, Cesium potassium antimonide (CsK 2 Sb) photocathode is of interest because of its low emittance, excitability in visible light, and high quantum efficiency (QE). The challenge is its high susceptibility to environment that lead to low operating vacuum pressure and short lifetime/low extraction charge. To resolve these issues, it is necessary to understand the molecular structure of the cathode and its degradation mechanism. In this study, we transported CsK 2 Sb photocathode to a beamline of synchrotron radiation facility using a vacuum transport system for molecular structure analysis. Specifically, the cathode was deposited in an evaporation system at Nagoya University. We transported it to Aichi Synchrotron Radiation Center (Aichi SR) located 15 km away, and analyzed it in the depth direction by X-ray photoelectron spectroscopy (XPS) at BL7U. Based on the results, we quantitatively evaluated the composition ratios and stoichiometry of the cathode elements (Sb, K, Cs). A Cs ex-cess state of surface was observed at the surface, and it is consistent with previous studies. The intended atomic structure of CsK 2 Sb was formed only at a few nanometres of the surface on the Mo substrate. On the other hand, the CsK 2 Sb cathode structure on the graphene substrate was preserved further in the depth direction.

47 OTHER INSTRUMENTATION↗
DOE OSTI2024

Probing thermodynamics of radiogenic helium and defects in $δ$-plutonium alloys and interactions with adsorbed environmental gases

Differential scanning calorimetry coupled with simultaneous evolved gas analysis (DSC-EGA) on aged δ-Pu samples shows that most radiogenic helium remains trapped within the Pu matrix at temperatures very close to, or slightly above, the melting temperature. Our results indicate that helium release from 50-year-old δ-Pu occurs as a burst just below the melting temperature (>0.994 T m ), with subsequent pressure oscillations as temperature increases. Subordinate quantities of H 2 were also released along with helium. Here, the helium emission tails off and ceases above ~720 – 750°C. In a δ-Pu alloy aged 6 years, the initial helium burst occurs slightly above melting (~1.015 – 1.042 T m ), with a discrete, larger helium spike occurring between 670 and 686°C. The proximity of helium release to the liquidus transition presented challenges in the deconvolution of overlapping process enthalpies, the liquidus endotherm, and the exotherm resulting from bubble collapse, annealing and gas expulsion. Helium’s strong affinity for vacancy binding in a 2He-vac configuration is predicted by Density Functional Theory (DFT) modeling. The measured stored energy associated with the He release events in a 50-year-old δ alloy is on the order of ~ 10–11 J/g, which is significantly higher than stored energies measured in the sub-solidus regimes (~2 J/g) that are related to the solid-state annealing of processing- and radiation-induced defects. This implies that aged δ Pu alloys have a remarkable resilience to accommodate the lattice strain produced by the internal pressure of the helium bubbles and provides further insight into the thermodynamic behavior of aged δ Pu.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Milestone 1.2.12: The Fate of Oxygen Radical Species in Corroded Aluminum Alloys under Irradiation

The U.S. Department of Energy tasked Idaho National Laboratory to evaluate the feasibility of extended dry storage of aluminum-clad spent nuclear fuel (ASNF) in heliumbackfilled canisters. A significant research effort has been devoted to determining the amount of radiation-induced molecular hydrogen (H 2 ) generation from corrosion layers that would be present on the fuel assembly surfaces. However, limited attention has been directed to the oxygen radical species that are concurrently generated in these H 2 producing radiation environments. This report collates and summarizes the available experimental and computational studies on the fate of these reactive oxygen radicals during the irradiation of aluminum oxyhydroxide and hydroxide polymorphs typically formed on ASNF. Based on this review, it is determined that radiation-induced oxygen-centered radicals do not react to give molecular oxygen gas (O 2 ), but rather remain trapped in the aluminum corrosion layers in their original crystal lattice positions. Overall, O 2 generation will not be a concern for extended storage of ASNF in helium-backfilled canisters. That is, no O 2 contribution to pressurization of sealed dry storage systems or creation of a flammable atmosphere is anticipated.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2020

Measuring aerosol collection efficiency for the Bladewerx “New Speclon TM 5” and the older “Speclon TM 5” filter

Bladewerx TM LLC (Rio Rancho, NM) manufactures instrumentation, neutron shielding and activation foils for the radiation protection industry. Specializing in portable alpha/beta air monitors and sample counters, Bladewerx is the source of Speclon TM PTFE filter media that they recommend for high-resolution alpha spectroscopy. Los Alamos National Laboratory (LANL) utilizes Speclon TM filter material in CAM (Continuous Air Monitor) samplers for workplace air monitoring. The LANL Aerosol Engineering Facility received air filter material from Bladewerx, referred to as “New Speclon 5” in this document, in order to distinguish from filter material that was previously received (referred to as “Speclon 5” in this document). In this document, the aerosol collection efficiency and airflow resistance (pressure drop) were measured for the New Speclon 5 filter material.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2024

Bounding Pressure and Flammability Evaluations of Aluminum-Clad Spent Nuclear Fuel Department of Energy Standard Canisters

This paper presents bounding pressure and flammability evaluations for DOE Standard Canister loaded with DOE-managed aluminum-clad spent nuclear fuel (ASNF). The objective of these evaluations is to gain confidence in the safety and feasibility of possible loading configurations for extended (>50 years) periods of dry storage, with particular focus on the dry storage canister pressures and potential for formation of a flammable atmosphere. The primary concern about the extended dry storage of ASNF is radiolytic gas generation. The aluminum cladding of these materials tends to corrode, and these corrosion products—typically aluminum oxides, such as boehmite, bayerite, or gibbsite—could carry water. This makes ASNF dry storage canisters difficult to dry. The gamma radiation field in dry storage environments could cause a radiolytic breakdown of residual water, forming chemical species such as molecular hydrogen (H2). The release of these species could increase the canister pressure and lead to the generation of a flammable canister atmosphere. The bounding evaluations presented within this study surmise conservative, but credible, conditions and processes. This includes the assumption of a full breakdown of a large quantity of free, physisorbed, and chemisorbed water (bound in a trihydrate, i.e., Al2O3 • 3H2O, layer). The considered dry storage configurations include a ~3 m (10 ft) long, ~46 cm (18 in) diameter (10 x 18) DOE Standard Canister loaded with 32 Advanced Test Reactor (ATR) ASNF elements, and a ~3 m (10 ft) long, ~61 cm (24 in) (10 x 24) diameter DOE Standard Canister loaded with 40 ATR ASNF elements. While the results of this study indicate the possibility of atmospheric hydrogen concentrations above the lower flammability limit, insufficient concentrations of oxygen will prevent the formation of flammable atmospheres. The maximum credible pressures remain well within the structural limits of the DOE Standard Canister.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Spatiotemporal measurements of striations in a glow discharge’s positive column using laser-collisional induced fluorescence

Here we have observed the behavior of striations caused by ionization waves propagating in low-pressure helium DC discharges using the non-invasive laser-collision induced fluorescence (LCIF) diagnostic. To achieve this, we developed an analytic fit of collisional radiative model (CRM) predictions to interpret the LCIF data and recover quantitative two-dimensional spatial maps of the electron density, n e , and the ratios of LCIF emission states that can be correlated with T e with the use of accurate distribution functions at localized positions within striated helium discharges at 500 mTorr, 750 mTorr, and 1 Torr. To our knowledge, these are the first spatiotemporal, laser-based, experimental measurements of n e in DC striations. The n e and 447:588 ratio distributions align closely with striation theory. Constriction of the positive column appears to occur with decreased gas pressure, as shown by the radial n e distribution. We identify a transition from a slow ionization wave to a fast ionization wave between 750 mTorr and 1 Torr. These experiments validate our analytic fit of n e , allowing the implementation of an LCIF diagnostic in helium without the need to develop a CRM.

42 ENGINEERING↗
DOE OSTI2024

Multi-physics Modeling of Radiative Heat Transfer and Fluid Flow for the Reactor Cavity Cooling System

High-temperature gas-cooled reactors (HTGRs) are notable for their high thermal efficiency and potential for combined heat and power applications. These reactors are particularly appealing due to their advanced passive safety features. HTGRs utilize passive safety systems that function without requiring active components like pumps or compressors during emergencies. These reactor designs depend on a Reactor Cavity Cooling System (RCCS) to manage decay heat removal from the reactor pressure vessel (RPV) during accident conditions. The RCCS consists of vertical rectangular channels known as "risers" or riser ducts positioned around the RPV. These risers receive heat from the RPV through both convective and radiative heat transfer mechanisms. Understanding the interplay of multiple physical phenomena, such as fluid dynamics, heat transfer, and neutron interactions, is essential for the effective design and operation of nuclear reactors, particularly for systems like the RCCS. Multi-physics simulations provide a comprehensive approach to studying these interactions, offering detailed insights and enhancing accuracy. They are especially important in RCCS designs, where the interaction between radiative and convective heat transfer can significantly impact system performance. By leveraging multi-physics simulations, complex reactor behaviors can be modeled without compromising the fidelity of the underlying physical processes. This work aims to establish a robust methodology for coupling multiple physical processes in an air-cooled RCCS. By focusing on validating this multi-physics approach, the study involves designing test cases that simulate various conditions to verify the numerical models employed. The outcomes of this research will provide critical insights for accurately modeling and optimizing complex nuclear systems like the RCCS.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗