SEARCH · Engineering Papers
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.
The tritons sublibrary aims to describe nuclear reactions between incident triton particles and different nuclei. For ENDF/B-VIII.1, compared to ENDF/B-VIII.0, there were only evaluation updates for 4 He.
The competing 22 Ne(α, γ) 26 Mg and 22 Ne(α,n) 25 Mg reactions control the production of neutrons for the weak s process in massive and asymptotic giant branch (AGB) stars. In both systems, the ratio between the corresponding reaction rates strongly impacts the total neutron budget and strongly influences the final nucleosynthesis. A number of experimental studies have been performed over recent years which necessitate the reevaluation of the 26 Mg and 22 Ne(α, γ) 25 Mg reactions control the production 26 Mg and 22 Ne(α, n) 25 Mg reaction rates. Evaluations of the reaction rates following the collection of new nuclear data presently show differences of up to a factor of 500, resulting in considerable uncertainty in the resulting nucleosynthesis. To reevaluate the 22 Ne(α, γ) 26 Mg and 22 Ne(α,n) 25 Mg reaction rates using updated nuclear data from a number of sources including updating spin and parity assignments. With updated spin and parity assignments, the levels which can contribute to the reaction rates are identified. The reaction rates are computed using a Monte Carlo method which has been used for previous evaluations of the reaction rates in order to focus solely on the changes due to modified nuclear data. The evaluated 22 Ne(α, γ) 26 Mg reaction rate remains substantially similar to that of Longland et al. but, including recent results from Texas A&M, the 22 Ne(α, n) 25 Mg reaction rate is lower at a range of astrophysically important temperatures. Stellar models computed with newton and mesa predict decreased production of the weak branch s process due to the decreased efficiency of 22 Ne as a neutron source. Using the new reaction rates in the mesa model results in 96 Zr / 94 Zr and 135 Ba / 136 Ba ratios in much better agreement with the measured ratios from presolar SiC grains. The 22 Ne + α reaction rates 22 Ne(α, γ) 26 Mg and 22 Ne(α, n) 25 Mg have been recalculated based on more recent nuclear data. The 22 Ne(α, γ) 26 Mg reaction rate remains substantially unchanged since the previous evaluation but the 22 Ne(α, γ) 25 Mg reaction rate is substantially decreased due to updated nuclear data. This results in significant changes to the nucleosynthesis in the weak branch of the s process.
An architecture for an inertial confinement fusion system is disclosed. The system includes a fusion chamber for producing neutrons from a fusion reaction, and a laser system in which lasers are arranged about a vacuum chamber to provide energy to the fusion chamber to initiate the fusion reaction. The beam paths between the lasers and the fusion chamber are configured to prevent neutrons from the fusion chamber from reaching the laser system at a level that would preclude human access to the laser system.
Here, this paper details the conceptual design of a thermal-to-14-MeV neutron converter consisting of a mixture of lithium and deuterium in a blanket material. Such a device operates a two-step reaction, first generating tritons via thermal neutron absorption in the tritium breeding material, and in the second step, high-energy neutrons are produced either via deuterium-tritium fusion reaction or with tritium reacting with lithium. A thermal-to-14-MeV neutron converter significantly hardens the neutron spectrum by virtually removing thermal neutrons and adding a high-energy 14 MeV component to the neutron spectrum. While similar concepts have been previously proposed and tested in other reactors, the unique characteristics of the Advanced Test Reactor (ATR), namely its important thermal flux (up to $10^{15}$ n.cm$^{-2}$.s$^{-1}$) makes it markedly attractive for obtaining a very large fast neutron flux, usable for irradiation studies under neutron flux conditions prototypical of fusion reactors. The paper provides a description of a new computational scheme developed for handling the coupled neutron-triton transport mechanism using the Geant4 toolkit. Resulting neutron spectra and high-energy neutron yields are summarized for different irradiation positions and potential neutron breeder materials. Maximum predicted thermal-to-14-MeV neutron yields are on the order of $2.10^{-4}$, which is consistent with previous studies found in the literature. Thus, when placed inside the ATR, such a neutron converter will be providing the largest high-energy neutron source available for activation and irradiation studies of materials foreseen for use in fusion reactors. Future steps will involve qualifying the computational scheme using the ATR critical facility using activation foil measurements.
This report presents the preliminary design (CED-1) of IER-552, pulsed-neutron die-away (PNDA) experiments with absorbers, to be conducted by Lawrence Livermore National Laboratory (LLNL). The goal of the PNDA experiments with absorbing materials is to produce high-quality, low-cost integral benchmarks that can be used to validate cross sections that are high priority nuclear data for the Department of Energy’s Nuclear Criticality Safety Program (NCSP). PNDA experiments can be used to generate nuclear data validation benchmarks quickly and cheaply, as they do not require fissile material or a nuclear facility to conduct them. The original PNDA experiments were set up at LLNL under IER-501. Extending the PNDA experiments to include absorbing materials leverages existing facilities and resources at LLNL to validate cross sections, with a basic experiment design that uses a neutron generator to impinge a short, mono-energetic neutron pulse on a poisoned, moderating target sample. The neutron pulse lasts 100 µs, after which the neutron population in the sample reaches thermal equilibrium and later spatial equilibrium. In this fully equilibrized state, the neutron population has a characteristic decay-time eigenvalue. The eigenvalue can be used as an integral parameter to validate nuclear data involved with neutron scattering and absorption. Two observables can be measured: the neutron population and the population of secondary γ-rays produced by inelastic and capture reactions. This report explores the feasibility of both approaches. For γ-ray measurements, this allows for direct measurement of reactions of interest for validation. During the neutron burst time, inelastic (prompt) γ-rays dominate, and these data can be isolated for validation of neutron inelastic scattering data. After the pulse ends, neutron capture (delayed) γ-rays dominate and are produced by many reaction channels as neutrons slow down to thermal energies. Two sets of γ-ray spectra (inelastic, capture) can be produced for validation. Additionally, the time-dependent decay of the capture γ-ray population when the neutron population is thermalized and in its fundamental mode can be used for validation. Two experiments can be imagined where pulsed neutrons are used to validate the nuclear data of absorbing materials. The first integrates well into the existing PNDA testbed at LLNL. The absorber material would be introduced into a moderating target and the same neutron generator, neutron detectors, electronics, and shielding box used for IER-501 would be employed. From the same experimental setup, many absorbing materials can be validated by simply changing the neutron poison in the moderator. The only significant modification would be the introduction of a γ-ray detector. Such an experiment would primarily serve to validate nuclear data in the thermal energy range. The second kind of experiment would validate the fast and epithermal energy range. It would involve pulsing neutron into a bare assembly of the absorber material and measuring the die-away of neutrons. The fast leakage of such an assembly requires nanosecond neutron-generator pulse widths and fast detectors/electronics that could measure ~200 ns die-away curves. This kind of experiment would ideally be performed at a cyclotron, like that available at Lawrence Berkeley National Laboratory.
Not Available
In arms control treaties, verification plays a crucial role in detecting non-compliance, deterring future violations, and building trust between state parties. Neutron interrogation that induces fission reactions in fissile isotopes and measures the resulting fission neutrons, could be employed for this purpose. This study aims to develop a system which can determine the presence of fissile material while intrinsically protecting information. In this paper, we focus on optimizing the system for discriminating between an enriched uranium block from a depleted uranium (DU) block. The system was built and we report on benchmark measurements with DU and 16% enriched uranium blocks. Furthermore, the Excalibur (Experiment for Calibration with Uranium) neutron source, a neutron spectrometer (redBubble Technology Industries (BTI) N-Probe), and superheated droplet detectors were used for these measurements. MCNP simulations provided insights into detector responses to fissile materials with varying isotopic compositions, confirming that the system functioned as designed.
Beryllium (Be), atomic number 4, is a silver gray metal of low density (1.85g/cm 3 ), moderately high melting point (1289°C), and quite good stability in the atmosphere. Favorable mechanical properties, particularly specific stiffness (elastic modulus/density), have resulted in a number of weight-critical structural applications. Other notable properties of beryllium include good dimensional stability, high specific heat and extremely high transparency to X-radiation. The nuclear properties of beryllium include a high neutron scattering cross section (6 barns for thermal neutrons) and a low neutron absorption cross section (0.009 barns for thermal neutrons) making it a good choice for a neutron reflector. The (n, 2n) reaction in beryllium also makes it suitable for a neutron multiplier.
We report fusion and fission reactions are profoundly dependent on hydrogen for sustained reactions. Fusion is fueled by the isotopes of hydrogen, and predominantly hydrogen-based moderators slow fission neutrons to propagate chain reactions. Intentional tritium production in fusion and concomitant tritium production in fission reactors introduce challenges. The technology to efficiently extract, harvest, and quantify tritium in and from advanced fission molten salt coolants and fusion molten tritium breeder materials will require more pronounced research and development. Measuring and quantifying hydrogen is necessary in all areas of nuclear materials. Although many characterization techniques cannot directly detect hydrogen, numerous techniques provide the necessary information to understand the behavior of hydrogen in nuclear materials.
In many practical applications, the accuracy of neutron transport calculations depends on an estimate of the energy distribution provided in the source definition. While the energy distribution of neutrons emitted by spontaneous fission sources is well-defined, the spectra of neutrons produced in (α,n) reactions in compound mixtures, such as americium-beryllium (AmBe), depend on the microstructural properties of the source material, which may vary. Although the energy of neutrons produced in nuclear reactions can be calculated from first principles, such computations require detailed knowledge of the key characteristics of the source material, which are often unknown. Here, this study focuses on deriving the at-birth neutron energy distribution for an AmBe source by applying reverse transport calculations to the externally measured high-precision spectrum recommended in the latest revision of the international standard. The derived at-birth spectrum was validated through indirect energy-sensitive neutron emission rate measurements of AmBe sources calibrated at national primary metrology institutions. The resulting at-birth neutron energy distribution serves as essential input data for radiation transport modeling, providing a validated spectrum for broader experimental and computational applications.
The 16 O($n, n'γ$) reaction was measured at the Los Alamos Neutron Science Center white neutron source using γ-ray detection in liquid scintillators present in the upper hemisphere of the Correlated Gamma-Neutron Array for sCattering (CoGNAC). Separate measurements of this reaction were performed using H 2 O and BeO targets in successive years. The unique high energies of γ rays emitted from the 16 O($n, n'γ$) reaction facilitated a clean selection of this reaction from threshold to 9.8 MeV incident neutron energy without the need for precise measurements of the γ-ray energy or the scattered neutrons. The precise time resolution of the liquid scintillator detectors was then exploited to obtain high-resolution incident neutron energy measurements, and good agreement was obtained between the H 2 O and BeO results reported here. The dominant literature data sets for this reaction have systematic differences between them, but the present results improve upon the neutron energy resolution of earlier measurements and show important discrepancies in recent data. Finally, tentative data are also shown up to 20 MeV incident neutron energy but are potentially subject to improved understanding of the relative γ-ray and α decay branches from 16 O excited states.
The response of LaBr 3 (Ce) and LaCl 3 (Ce) scintillators to fast neutrons is investigated. Neutron-induced charged-particle reactions are observed in both materials when exposed to the fast neutrons produced by an AmBe source, with pulse-shape discrimination used to separate channels. LaBr 3 (Ce) is found to have the best separation between reaction channels, while LaCl 3 (Ce) has a significantly higher efficiency.
Progress report for the period November 1970 through January 1971. Numerous short papers present progress at various labs.
Within an isospin- and momentum-dependent Boltzmann-Uehling-Uhlenbeck transport model, we investigate the hard photons emission from neutron-proton bremsstrahlung in the reaction system of 208 Pb + 208 Pb around Fermi energy. Effects of neutron skins in coordinate (r) space and proton skins in momentum (k) space on the time evolution, the angular distribution, and the transverse momentum spectra of hard photons with different energies are studied. It is shown that the emission of direct hard photons is sensitive to the neutron skin, which has larger effects for more energetic hard photons. Meanwhile, we find that the proton skins have an important influence on the emission of direct hard photons, and its effect is actually even larger than that of neutron skins. Furthermore, it needs to take the effect of proton skins into account when we determine the size of neutron skins by comparing transport mode predictions of hard photons with the corresponding experimental measurements.
The spin cutoff parameter for 59 Ni has been studied from different types of experimental data, including neutron angular distributions from the 56 Fe(α,n) 59 Ni reaction, spin of discrete levels, the level density from the proton evaporation spectrum of the 54 Fe( 6 Li,p) 59 Ni reaction and neutron resonance spacing. Experimental data points were compared with calculations using models widely used in literature. It was found that the available empirical models overestimate data points in the energy region below the neutron separation energy, while microscopic calculations which take into account pairing correlations within Hartree-Fock +Bardeen-Cooper-Schrieffer approach are consistent with data. As a result, it confirmed earlier findings that pairing correlations play an important role and need to be taken into account when the spin cutoff parameter is calculated below the neutron separation energy.
Two methods are examined for extending the life of tritium targets for production of 14 MeV neutrons by the 3 H( 2 H,n) 4 He nuclear reaction. With thick film targets the neutron production rate decreases with time due to isotope exchange of tritium in the film with implanted deuterium. Here, the target life is maximized by operating the target at elevated temperature where the implanted deuterium mixes by thermal diffusion throughout the entire thickness of the film. The number of neutrons obtained from a target is then proportional to the initial tritium content of the film. A novel thin-film target design was also developed and tested. With these thin-film targets, the incident deuterium is implanted through the tritide into the underlying substrate material. A thin permeation barrier layer between the tritide film and substrate, reduces the rate of tritium loss from the tritide film. Good thin-film target performance was achieved using W and Fe for the barrier and substrate materials respectively. Thin-film targets were fabricated and tested and shown to produce similar number of neutrons as thick-film targets while using only a small fraction of the amount of tritium.
The occupancies and vacancies of the valence neutron orbitals across the stable tin isotopic chain from 112 ≤ A ≤ 124 have been determined. These were inferred from the cross sections of neutron adding and -removing reactions. In each case, the reactions were chosen to have good angular momentum matching for transfer to the low- and high-` orbitals present in this valence space. Here, these new data are compared to older systematic studies. The effective single-neutron energies are determined by combining information from energy centroids determined from the adding and removing reactions. Two of the five orbitals are nearly degenerate, below N = 64, and approximately two MeV more bound than the other three, which are also degenerate.
The use of sensitivity and uncertainty-based techniques has been well-developed in the neutronic community for criticality safety applications. Sensitivity techniques calculate the first-order relative variations in keff due to relative cross-section variations by isotope, reaction type, and incident neutron energy. The result is a vector of sensitivity coefficients which is folded with prior cross-section uncertainties to calculate the so-called similarity indices. Similarity indices are integral quantities which serve as useful mathematical measures for assessing the relevance of existing or new experiments to given application conditions, representing design systems for which the models are to be validated. Despite their importance for a wide range of validation activities, the peer-reviewed literature is non-existent on the development of rigorous verification analysis for their calculated values. Thus, this manuscript proposes a simple non-intrusive method to calculate similarity indices, serving as an independent verification tool for their evaluation. The manuscript derives the theoretical basis for the proposed verification analysis. Next, the verification analysis is applied to the ORNL’s SCALE code suite which employs adjoint-based methods to calculate similarity indices. Numerical experiments using a number of well-known benchmark experiments are employed to verify the calculation of the similarity indices.