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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 1,243 records · Page 69
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Proof-of-concept of a neutron time-of-flight ellipsoidal detector

The time-resolved measurement of neutrons emitted from nuclear implosions at inertial confinement fusion facilities is used to characterize the fusing plasma. Several significant quantities are routinely measured by neutron time-of-flight (nToF) detectors in these experiments. Current nToF detectors use scintillators as well as solid-state Cherenkov radiators. The latter has an inherently faster time response and can provide a co-registered γ-ray measurement as well as improved precision in the bulk hot-spot velocity. This work discusses a nToF ellipsoidal detector that also utilizes a solid-state Cherenkov radiator. The detector has the potential to achieve a fast instrument response function allowing for characterization of the γ-ray burn history as well as the ability to field the detector closer to the fusion source. Proof-of-concept testing of the nToF ellipsoidal detector has been conducted at the National Ignition Facility using commercial optics. A time-resolved neutron signal has been measured from the diagnostic. Preliminary simulations corroborate the results.

Jeet, J. (ORCID:0000000291091563)↗
DOE OSTI2025

Evaluations for medium- and high-mass nuclei for FUSION applications

There is a renewed attention to nuclear fusion as a commercial source of carbon-free energy, however there are many scientific needs that must be addressed to enable the future success of fusion as an economical energy option. Among these is the proper description of the impact of radiation produced in the fusion vessel chamber and all other components of the reactor. In this work we will focus on the nuclear data needs to describe the interaction between primary and secondary neutron radiation and the medium- and high-mass nuclei commonly present in structural (such as stainless steel) and superconducting (e.g., electromagnets) materials.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Status of the RRR evaluations for 233,235 U and 239 Pu [Slides]

The presentation discusses the R -Matrix Analysis of 233 U, 235 U, and 239 Pu and how their interplay among PFNS, TNC, RRR, and neutron multiplicities has a significant impact on the benchmarks. The presentation also discusses the 233 U improved agreement on benchmarks, an additional investigation on the thickness of beryllium reflectors, work on URR (up to 40 keV) and/or extension of RRR (600 eV to 2 keV). Also discussed is how the 235 U progresses on understanding reactivity rates related to depletion calculations, improved agreement with measured (capture) data is quite insensitive to resolve the negative slope, and how the 239 Pu negative slope as a function of the temperature is not understood yet (bias over 2σ lower than measured).

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2021

The Effect of the Flux Separability Approximation on Multigroup Neutron Transport

The angular dependence of flux-weighted multigroup cross sections is commonly neglected when generating multigroup libraries. The error of this flux separability approximation is typically not isolated from other error sources due to a lack of availability of library generation and corresponding solvers that cannot relax this approximation. These errors can now be isolated and quantified with the availability of a multigroup Monte Carlo transport and multigroup library-generation capability in the OpenMC Monte Carlo transport code. This work will discuss relevant details of the OpenMC implementation, provide an example case useful for detailing the type of errors one can expect from making the flux separability approximation, and end with more realistic problems which show the impact of the approximation and highlight how it can strongly arise from an energy-dependent resonance absorption effect. Since the angle-dependence is intrinsically linked to the energy group structure, these examples also show that relaxing the flux separability approximation with angle-dependent cross sections could be used to reduce either the fine-tuning required to set a multigroup energy structure for a specific reactor type or the number of energy groups required to obtain a desired level of accuracy for a given problem. This trade-off could increase the costs of generating multigroup cross sections, and has the potential to require more memory for storing the multigroup library during the transport calculations, but it can significantly reduce the computational time required since the runtime of a discrete ordinates or method of characteristics neutron transport solver scales roughly linearly with the number of groups.

42 ENGINEERING↗
DOE OSTI2022

Update of the Nuclear Criticality Slide Rule: Review of the Estimation of the Number of Fissions

IRSN (France), LLNL (USA) and ORNL (USA) began a long-term collaboration effort in 2015 to update the nuclear criticality Slide Rule (https://ncsp.llnl.gov/analytical-methods/criticality-sliderule) for the emergency response to a nuclear criticality accident. The Slide Rule permits the estimation of neutron and gamma dose rates and integrated doses based upon estimated fission yields, as a function of distance from the fission source, and time after criticality accidents for different critical systems. This paper presents the review of the section “estimation of the number of fissions”, named “fission yield” in the Slide Rule document. This estimation is important because the other information provided by the Slide Rule is directly proportional to the number of fissions. The previous models, based on Hansen and Barbry formulae, will be presented with their associated assumptions and limitations. Then, new proposals will be presented with a comparison to past criticality accidents and experimental data. The new formulae, based on the heat energy equation, cover many types of criticality accidents (solution, dry or low-moderated powder, rods/assemblies in water and dry metal) and require limited information in an emergency situation.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Update of the Nuclear Criticality Slide Rule Calculations: Plutonium systems – Delayed Fission Gamma

IRSN (France), LLNL (USA) and ORNL (USA) began a long-term collaboration effort in 2015 to update the nuclear criticality Slide Rule for the emergency response to a nuclear criticality accident. The Slide Rule permits the estimation of neutron and gamma dose rates and integrated doses based upon estimated fission yields, as a function of distance from the fission source, and time after criticality accidents for different critical systems. This paper presents results from the fourth phase of the current update of the Slide Rule project, in which delayed fission-product gamma (DFG) dose rates of unreflected plutonium critical systems were compared by several modern 3D radiation transport codes (MCNP, COG, SCALE), using updated flux-to-dose conversion factors. Dose rates are calculated for fissile material at five moderation ratios (H/Pu), at 1 m above the ground as a function of distance (between 30 cm and 1,200 m) from the external surface of the source to the center of the detector, and for periods between 1 s and 1,000 min after the critical instantaneous event. Further efforts have been devoted to the determination of the delayed gamma source, by comparing the time-dependent energy spectra obtained from several methods. Overall, DFG dose rates calculated by each participant led to consistent results. Extra effort is under way to identify the cause of the remaining differences, by comparing precisely the gamma source, and particularly nuclides inventories.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

MCNPs Easy Sources for (α,n) (MESA) 1.0: A User’s Guide

MCNPs Easy Sources for Alpha,N (MESA) facilitates the construction of MCNP source definition cards for (α,n) sources. It is a complete re-write of the homogeneous, surface source, and three layer problem types from Sources4c. Please refer to this reference for documentation on the cross-sections and neutron emission spectra. The one departure from Sources4c is the use of alpha energies and activities from the ISC ENDF VII or VIII library specified by the user.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

The CLAS12 Backward Angle Neutron Detector (BAND)

The Backward Angle Neutron Detector (BAND) of CLAS12 detects neutrons emitted at backward angles of 155° to 175°, with momenta between 200 and 600 MeV$/c_0$. It is positioned 3-m upstream of the target, consists of 18 rows and 5 layers of 7.2-cm by 7.2-cm scintillator bars, and read out on both ends by PMTs to measure time and energy deposition in the scintillator layers. Between the target and BAND there is a 2-cm thick lead wall followed by a 2-cm veto layer to suppress gammas and reject charged particles. This paper discusses the component-selection tests and the detector assembly. Timing calibrations (including offsets and time-walk) were performed using a novel pulsed-laser calibration system, resulting in time resolutions better than 250 ps (150 ps) for energy depositions above 2 MeV (5 MeV). Cosmic rays and a variety of radioactive sources were used to calibration the energy response of the detector. Scintillator bar attenuation lengths were measured. We report the time resolution results in a neutron momentum reconstruction resolution, $δp/p$ < 1.5% for neutron momentum 200 ≤ $p$ ≤ 600 MeV/c. Final performance of the BAND with CLAS12 is shown, including electron–neutral particle timing spectra and a discussion of the off-time neutral contamination as a function of energy deposition threshold.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2021

Feasibility of full-core pin resolved CFD simulations of small modular reactor with momentum sources

Complex flow structure interactions and heat transfer processes take place in nuclear reactor cores. Given the extreme pressure/temperature and radioactive conditions inside the core, numerical simulations offer an attractive and sometimes more feasible approach to study the related flow and heat transfer phenomena in addition to the experiments. Under the Exascale Computing Project, the full-core simulation of a small modular reactor (SMR) has been pursued coupling Computational Fluid Dynamics (CFD) and neutronics. A key aspect of the modeling of SMR fuel assemblies is the presence of spacer grids and the mixing promoted by mixing vanes or the equivalent. A reduced order methodology is adopted based on momentum sources to mimic the mixing of the vanes. The momentum sources have been carefully calibrated with detailed Large Eddy Simulations (LES) of spacer grids performed with Nek5000. Modeling the spacer grid and mixing vanes (SGMV) effect without body-fitted computational grid avoids the excessive costs in resolving the local geometric details, and thus supports the simulation to be scaled up to the full core. Besides the progress on momentum source modeling, this paper also features the first full-core pin resolved CFD simulation ever performed to the authors' knowledge. This represents a significant advancement in capability for the CFD of nuclear reactors, which will hopefully serve as an inspiration for further integrating high-fidelity numerical simulations in actual engineering designs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Errors introduced in fission neutron spectrum measurements using a single reference

Measurements of physical quantities are often made relative to a reference quantity, simplifying the interpretation of the measurement and its uncertainties. However, if applied without careful considerations, this technique can carry with it a significant and complex systematic error that is usually ignored. For example, measurements of continuous neutron spectra relative to a reference spectrum yield a result that does not properly include the important effects of neutron scattering in the experimental environment. These effects act to distort the measured spectrum shape and create notable errors in spectrum-integrated quantities. While this effect is one of a series of known potential sources of systematic error including, but not limited to target impurities, angular distributions, incident beam flux, and others, the error introduced from differences in environmental scattering effects between the reference and desired spectra are frequently overlooked. In this work, we demonstrate the origin of this error using simulated data corresponding to measurements of the neutron-induced prompt fission neutron spectra of 233 U, 235 U, 238 U, and 239 Pu as well as the spontaneous fission neutron spectrum of 252 Cf. Errors in the spectrum shape can be as high as 10%–15% depending on the combination of incident and outgoing neutron energy. The spectrum integral and average spectrum energy also contain errors that change with incident neutron energy and can be as high as ~2%. Although this study is based on simulated data, we also show a method for proving that this error exists in an experimental data set and we show a correction that can be applied to results of this type to obtain a more accurate result, thereby largely removing systematic errors from environmental scattering effects. The potential for these kinds of errors should be considered in measurements producing continuous neutron spectra.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2025

New class of tritium breeders for fusion applications: Metal-reinforced composite breeders

Commercial fusion reactors operating on a D-T fuel cycle will require a steady supply of tritium to maintain the burning plasma required for continuous power generation. Since tritium has a short half-life, there is negligible natural abundance which necessitates fusion reactors to produce their own source of tritium. Tritium is most easily produced by surrounding a fusion reactor core with lithium (Li), which reacts under the intense neutron flux leaving the reactor core to form tritium and helium. Due to the hazards and technical challenges associated with surrounding a fusion reactor core with many tons of molten Li, other Li-bearing tritium breeder materials have been pursued. Unfortunately, most of the liquid breeders historically examined are exceedingly corrosive to reactor structural materials while many solid breeders in the form of ceramics are forced to make tradeoffs between Li content and mechanical integrity. In this work, to break the historic limit between Li-density and mechanical integrity of traditional solid breeders, a new class of solid tritium breeders is developed: metal-reinforced composite (MERC) breeders. Specifically, the high Li-density of lithium oxide (Li 2 O) is exploited through the addition of a metal reinforcing phase, producing a composite breeder material exhibiting high splitting tensile strength and quasi-ductility with a Li-density greater than other leading solid breeder candidates, including lithium orthosilicate (Li 4 SiO 4 ) and lithium metatitanate (Li 2 TiO 3 ). Mechanical testing, microstructural characterization, and neutronic simulation results are presented and discussed in light of fusion reactor design considerations.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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 OSTI2021

Application of Shift Ex-Core Calculations to a Detailed MSR Model

This report documents the collaborative research conducted between Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL) on applying the Monte Carlo neutronics code Shift to calculate the radiation dose rates for a detailed molten salt reactor (MSR) model based on the Molten Salt Breeder Reactor (MSBR). In this work scope, the ANL and ORNL teams worked together and applied Shift to calculate the radiological environments of this detailed MSR facility modeled with explicit geometries. The radiological conditions within the MSR facility were modeled for when the reactor is at two different operation modes: normal full power and drained state. The FW-CADIS hybrid method in Shift was applied successfully to calculate the ex-core neutron and gamma dose rates for the MSBR at full power. Dose rate maps showed that in the current MSBR numerical model, potential pathways exist for neutrons and gammas to stream through the 8-foot concrete shield to reach the top of the MSBR reactor cell. Neutron and gamma dose rates within the drain cell were also calculated for the MSBR at the drained state by integrating the source terms obtained from an ORIGEN-S depletion calculation into the Shift simulation. The results indicate that in the drained state, the delayed gammas from the depleted fuel salt are the main contributors to the dose rates, which suggests that a steel liner in the tank model must be added for calculating the dose rates in the drain cell.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

A summary of air sampler filter performance data: Bladewerx Speclon 5(TM) and Millipore FMLW047 5um

Bladewerx™ 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 5™ PTFE filter media that they recommend for high-resolution alpha spectroscopy. Los Alamos National Laboratory (LANL) utilizes Speclon™ filter material in CAM (Continuous Air Monitor) samplers for workplace air monitoring. Millipore FMLW (5μm) filters are also used at Los Alamos, and a comparison of the two filter types has been requested. For filter face air velocities from 0.066 m/s to 1.5 m/s, the aerosol collection efficiency (for 0.3 μm particle diameters) and the filter pressure drops were measured. The FWHM (full width half maximum) of alpha spectroscopy peaks was also determined, using naturally occurring radon progeny at an alpha energy of 6 MeV.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2025

Q3 Report for FY25 Theory and Simulation Performance Target: Development of an integrated modeling framework for fusion reactor design and assessment

This report describes the work and activities carried out towards the completion of each of the following milestones in FY25 Q3: 1. Demonstrate workflow for generating self-consistent CESOL plasma profiles + first wall and divertor loading prediction and generate the CAT plasma and neutron loading needed for further engineering analysis. • Benchmark between two first wall heat flux mapping methods, identify importance of various heat flux sources and physics impact of using fully coupled CESOL vs post-analysis evaluation. 2. Generate medium fidelity parametrized CAD. • Generate parametrized CAD components for the CAT example case via either user-defined modules called within the geometry generation or by defeatured/parametrized CAD, including DCLL blanket matched to divertor boundary and magnets. Define materials, labels, and boundary conditions for passing the mesh to CFD tools. 3. Demonstrate multiphysics magnet analysis. • Demonstrate magnet analysis workflow called from the FREDA workflow, and 4. Demonstrate nuclear analysis. • Add model to OpenFOAM and/or other codes possibly including Diablo to account for tritium diffusion in solids.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Calculating Radiation Damage (DPA) from Transmutation Products

This is a poster for an INL poster session. Accurate models for radiation damage are crucial for predicting material performance in radiation environments. The uncertainty of state-of-the-art radiation damage models is large, contributing to excessive safety margins. A major source of this uncertainty is neglecting the effect that transmutation products have on radiation damage. Transmutation products are new nuclides formed by neutron activation during irradiation; they can contribute to radiation damage by additional neutron capture or decay events. Ignoring the contribution of transmutation products leads to a significant underprediction of the radiation damage (e.g., >10% error in 316 stainless steel). This underprediction is accounted for in part by adding larger safety margins to designs. Currently, the state of the art explicitly accounts for only a single transmutation product, namely nickel-59, during the radiation damage calculation. All other transmutation products are assumed to not contribute to the radiation damage, because there is currently no established method to systematically track all or a selection of radiation damage contributions of transmutation products during activation. In the case of nickel-59, the current method is to apply a precalculated correlation that cannot be used for any other nuclide and is largely dependent on all nuclear engineers being experts in this niche topic. This project proposed to methodically find other transmutation products that cause significant radiation damage, and then to develop a general framework for systematically tracking the radiation damage from these nuclides. This was accomplished by combining the radiation damage calculation into the transmutation calculation already performed for irradiated structural materials. The key idea of our framework is to introduce radiation-damage "pseudo-nuclides" to the list of nuclides used in the transmutation analysis. This allows radiation damage to be tracked alongside the creation and destruction of transmutation products. The main deliverable of this project is a general framework for computing radiation damage while the damaged material undergoes transmutation; this capability allows a significantly more accurate estimation of radiation damage, and in turn reduce required safety margins thereby reducing the cost to construct reactors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Neutron response of the TLYC scintillator

The response of the recently developed Tl 2 LiYCl 6 (Ce) scintillator to neutrons was measured and the 35 Cl(n,p) channel observed in the material for the first time. The scintillator was exposed to a 252 Cf source mounted within a parallel-plate avalanche counter for fission-fragment coincident measurements. Proton- and -like events are selected with the pulse-shape discrimination technique. Additionally, comparison of energy-deposition and time-of-flight allows the quenching of the different channels to be assessed. The 6 Li(n,t) channel is found to have its energy quenched to 37(1)% relative to -ray detection, while the proton energy deposition spectrum is quenched to 60(1)%.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗