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At least 307 records · Page 17
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

System and method for fissionable material detection with a short pulse neutron source

A system and method is disclosed for detecting fissionable materials. In one embodiment the system may incorporate a neutron pulse generator configured to generate multiple short pulses of neutrons, or a single pulse of sufficient intensity, in a vicinity of an object of interest. The source pulse of neutrons includes neutrons which each have a full width half maximum time duration of less than about 100 ns and a peak energy level no greater than about 20 MeV. A fast response detector is used which is able to detect single neutron events indicative of fission neutrons having been produced by the source pulse of neutrons interacting with fissionable material associated with the object of interest, and which arrive at the fast response detector within a predetermined time window immediately before arrival of the source neutron pulses.

Podpaly, Yuri↗
DOE OSTI2024

Modernizing the Legacy Fission Wire Measurement System for the Advanced Test Reactor-Critical Facility

Operational lifetime extensions of existing research reactors have emphasized the need for refurbishment, replacements, and upgrades to supporting equipment and instrumentation. The Advanced Test Reactor (ATR) at Idaho National Laboratory (INL), which entered service in 1967, has recently completed the sixth core internals change-out and has scheduled operations until at least 2040. Reactor maintenance and operational risk management is critically important in the research reactor community, however supporting measurement systems sometimes get overlooked when maintenance is planned. The Fission Wire Measurement System (FWMS) is a custom measurement system designed in the 1960s to measure the beta-particle activity of irradiated uranium-aluminum fission wires. This measurement is conducted to determine the fission rate profile of the Advanced Reactor Test Critical (ATR-C) facility. The ATR-C is an open-pool, low-power test reactor that was purpose driven to resemble ATR and is used to qualify experiment configurations and verify core models prior to full-power experiment irradiations in ATR. A power distribution measurement in ATR-C uses uranium-aluminum wires that are distributed throughout the ATR-C core to validate simulation and modeling results. These measurements require 340 to 1500 wires to be irradiated and measured within a 12-hour window. The activity of the wires is measured in the required time with the FWMS, which was put into service in 1965 at the Radiation Measurements Laboratory (RML). The system consists of 4 measurement channels and one reference channel, each with a 2-pi proportional gas flow detector and the measurement channels each have an automated sample changer. This legacy system is crucial to the continued operations of ATR and has undergone some minor hardware upgrades since 1965, however the system presently relies on custom control boards, custom gas ion chambers, analog amplifiers/discriminators, and a user interface (UI) for the system written in outdated code. Much of the equipment and software is custom with no commercial replacements or support and limited documentation. The existing control software requires an operating system that is no longer supported, creating more vulnerabilities to continued operations. A project is underway with a third-party vendor to design, build, and document a new control and data acquisition system (CDAS) for the FWMS. The new upgrade will replace the control system, computer, UI, sample changer motors, and main power supply while maintaining the interface with existing detector hardware. The upgraded system will be operated in parallel with the current hardware and software to conduct validation testing. This equipment upgrade demonstrates the commitment at ATR to ensuring successful operations and potential future research reactors at INL.

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗
DOE OSTI2020

Consistent Evaluation of the Prompt-fission Neutron Spectrum and Multiplicity for n+ 235,238 U and n+ 239 Pu

This report was written to satisfy a FY20 NCSP milestone on 235,238 U and 239 Pu. The milestone requires to “finalize a report assessing our methodology to evaluate prompt-fission neutron spectrum (PFNS) and multiplicity consistently”. More specifically, we study whether the code CGMF can reproduce ENDF/B-VIII.0 evaluated PFNS and average prompt-fission neutron multiplicities, $\bar{v}$, for 235,238 U and 239 Pu using one joint parameter set per isotope. If CGMF is shown to be able to reasonably reproduce ENDF/B-VIII.0 within its model-parameter space, this code could be used for future consistent evaluations of PFNS and $\bar{v}$. To answer this question, we explore here the parameter space of CGMF and its impact on calculated values and whether they are close to evaluated and experimental data. We also list experimental data that would enter a future evaluation and statistics method that could be used to obtain evaluated data and covariances. We conclude that values of $\bar{v}$ calculated by CGMF are reasonably close to ENDF/B-VIII.0 data, while more work on modeling the PFNS is needed (parameter optimization and model improvements) to reliably use it for evaluations.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2022

R-Value Measurements Performed on Uranium Targets Irradiated with Fission Spectrum Neutrons FY 2021 for F2019 Project

The separation and characterization of two irradiated uranium targets, a depleted uranium (DU) and a highly enriched uranium (HEU) target, was conducted in April of 2021. The two targets were assembled at Los Alamos National Laboratory (LANL) and irradiated using the critical assembly at the National Criticality Experiments Research Center (NCERC). Splits of the dissolved targets were received by Pacific Northwest National Laboratory (PNNL) after which the PNNL and LANL teams chemically separated the solutions using independent separation schemes and analyzed the separated fractions for short lived actinides and fission products. Chemical separations at PNNL were traced with stable or radioactive tracers to allow for the determination of chemical yields, analyzing using either inductively coupled plasma optical emission spectroscopy (ICP-OES), inductively coupled plasma mass spectrometry (ICP-MS) or gamma emission analysis (GEA) depending on the nature of the tracer. Several other analytical techniques were used by PNNL including kinetic phosphorescence analysis (KPA) and thermal ionization mass spectrometry (TIMS) depending on the analyte’s need. Comparisons were made between current and historical PNNL and LANL, as well as literature values. Overall, there was agreement between the two laboratories for the bulk of analytes, with some notable exceptions such as 111 Ag, and 141,143,144 Ce. Included in these comparisons were the short-lived actinides 237 U, 239 Np, the fission products 89 Sr, 91 Y, 95/97 Zr, 99 Mo, 111 Ag, 115/115 mCd, 136/137 Cs, 140 Ba, 141/143/144 Ce, 147 Nd, 153 Sm, 156 Eu, and 161 Tb, providing both total atoms as well as the R-values. The data presented in this report represents the sixth NCERC irradiation of HEU and DU (FY13, FY14, FY15, FY17, FY18, FY21), and their subsequent separation and analysis.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Electronic couplings for singlet fission: Orbital choice and extrapolation to the complete basis set limit

Abstract For the search for promising singlet fission candidates, the calculation of the effective electronic coupling, which is required to estimate the singlet fission rate between the initially excited state (S 0 S 1 ) and the multiexcitonic state ( 1 TT, two triplets on neighboring molecules, coupled into a singlet), should be sufficiently reliable and fast enough to explore the configuration space. We propose here to modify the calculation of the effective electronic coupling using a nonorthogonal configuration interaction approach by: (a) using only one set of orbitals, optimized for the triplet state of the molecules, to describe all molecular electronic states, and (b) only taking the leading configurations into consideration. Furthermore, we also studied the basis set convergence of the electronic coupling, and we found, by comparison to the complete basis set limit obtained using the cc‐pV n Z series of basis sets, that both the aug‐cc‐pVDZ and 6–311++G** basis sets are a good compromise between accuracy and computational feasibility. The proposed approach enables future work on larger clusters of molecules than dimers.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2023

Fusion Innovation and the Valley of Death—Lessons from Fission

Over the past several decades, the United States (U.S.) and other countries have done important research and development in advanced nuclear energy technologies, both fission and fusion. The urgency of our current climate, energy, and security needs have prompted a concerted effort by innovators, investors, and the public sector to transition from research and development (R&D) to demonstration and commercialization—to cross the proverbial “valley of death.” The stakes are high: humanity needs clean energy options, and the failure of any promising alternative will lower our odds of success. The process of moving an innovation into use has often been described in phases that begin with early research and end with deployment at scale. Between those two phases lies an area often referred to as the “valley of death” because it can require a long period of time, large investments, and often includes significant challenges related to regulation, standardization, and entering markets already structured around incumbent technologies. What can we learn from recent experience with fission to inform our path forward with fusion? At least five key lessons rise to the top: 1) Set bold targets; 2) Empower the private sector; 3) Engage with regulators and the public; 4) Address costs and markets; and 5) Deliver results. Here, this editorial will consider each of these in turn.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Improvement of the $\mathrm{BISON U_3Si_2}$ modeling capabilities based on multiscale developments to modeling fission gas behavior

Uranium silicide (U 3 Si 2 ) is a concept explored as a potential alternative to UO 2 fuel used in light water reactors (LWRs) since it may improve accident tolerance and economics due to its higher thermal conductivity and increased uranium density. U 3 Si 2 has been previously used in research reactors in the form of dispersion fuel, but operated at lower temperatures than commercial LWRs. The research reactor data illustrated that significant gaseous swelling occurs as the fuel burnup increases. Therefore, it is imperative to understand the fission gas behavior of U 3 Si 2 under higher temperature LWR operating conditions. In this work, molecular dynamics and phase-field modeling techniques are used to reduce the uncertainty in select modeling assumptions made in developing the fission gas behavior model for U 3 Si 2 in the BISON fuel performance code. These lower length scale informed models are then utilized in the validation of BISON U 3 Si 2 modeling capabilities to simulate the ATF-1 experiments irradiated in the Advanced Test Reactor (ATR). Sensitivity analysis (SA) and uncertainty quantification (UQ) are included as part of the validation process to identify where further experiments and lower length scale modeling would be beneficial. Here, the multiscale modeling approach utilized in this work can be applied to new fuel concepts being explored for both LWRs and advanced reactors (e.g., uranium nitride, uranium carbide).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Verification of Bison fission product species conservation under TRISO reactor conditions

When assessing the reliability and predictive capabilities of a simulation tool, code verification is used to ensure that the implemented numerical algorithm is a faithful representation of its underlying mathematical model, including partial differential or integral equations, initial and boundary conditions, and auxiliary relationships. During this process, numerical results in a discrete solution are compared to the analytical solution of the mathematical model. Here, in this paper, the code verification process is applied to one-dimensional spatiotemporal problems that exercise partial differential equation governing the conservation of fission product species (or mass diffusion). Numerical experiments were performed in the Bison fuel performance code to evaluate its predictive capability under various TRISO reactor conditions such as base irradiation and safety heating test conditions for either short- or long-lived fission product species, as well as a case concerning evaporation from the outer surface of a particle. The code predictions were compared with the expected exact results obtained from the analytical expressions, and the fact that they demonstrate the correct analytical behavior provides strong evidence of proper numerical algorithm implementation.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2023

Catalyzed oxidation of nuclear graphite by simulated fission products Sr, Eu, and I

The influence of three fission products Sr, Eu, and I on the oxidation of IG-110 nuclear graphite was studied in the temperature range of 400 to 1000 °C. Sr and Eu were introduced as chlorides, and I was introduced as NaI. The temperature dependence of both CO 2 and CO production during the graphite oxidation measured with mass spectroscopy and infrared spectrometry shows that the introduction of these three compounds to graphite significantly decreases the onset temperature for the oxidation of graphite. Among the three compounds, NaI is the most active towards the oxidation reaction, characterized by a significant decrease of the onset temperature from approximately 650 to 400 °C before and after its introduction to graphite. Separate measurements of CO 2 and CO concentration at varying temperatures enable the calculation of the activation energy for the formation of CO 2 and CO. The activation energies for the oxidation of pure and fission product-impregnated graphite samples decrease in the following order: standard IG-110 graphite, EuCl 3 -impregnated IG-110, SrCl 2 -impregnated IG-110, and NaI-impregnated IG-110. This trend indicates that the three compounds catalyze the oxidation of graphite at temperatures relevant to the operation of high-temperature gas-cooled reactors. Furthermore, it is found that the three compounds can also affect the molar ratio of reaction products CO 2 and CO, and the rates of the graphite oxidation. At temperatures higher than about 850 °C, the impregnated samples exhibit lower CO 2 : CO ratios than the pure graphite. Different from EuCl 3 and NaI, the introduction of SrCl 2 decreases the graphite oxidation rates at temperatures higher than about 770 °C. Their catalytic mechanism can be understood based on a redox cycle of the intermediate active species, promoting the dissociation of molecular oxygen and transfer to the carbon.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Single-size and cluster dynamics modeling of intra-granular fission gas bubbles in UO 2

For this work, we perform simulations of intra-granular fission gas bubble evolution in UO 2 using both a relatively simple, computationally inexpensive single-size model and a detailed cluster dynamics model. Simulations encompass 36 experimental cases from 4 different databases, covering various temperature and burnup levels. We systematically compare results from the two models to each other and to post-irradiation experimental data of bubble average size and number density. Overall, the model-to-model comparisons reveal an excellent agreement across the set of simulations. This outcome indicates that, in spite of the underlying assumptions, the single-size model provides a good approximation of the complex physical behavior that is more rigorously described by the cluster dynamics model. Qualitatively, both models reproduce the trends of the experimental data with temperature and burnup correctly. Quantitatively, calculated results are either in good agreement with the data or within errors that appear consistent with the inherent uncertainties. Moreover, for the single-size model, we demonstrate and assess a multiscale approach whereby values for the fission gas atom diffusion coefficient from separate atomistic calculations are used. Systematic comparisons to experimental data point out a credible accuracy of the multiscale model.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Short Communication: Observation of Initial Burst Release of Fission Gas from High-Burnup UO 2 Nuclear Fuel During Thermal Transient

A system was developed and tested to provide a deeper understanding of the fission gas release kinetics from nuclear fuel during thermal transients. Pressure, temperature, spectral data, and optical images were simultaneously collected during resistive sample heating, and all released gases were collected in liquid nitrogen-cooled charcoal traps. Standup testing was performed with a high-burnup nuclear fuel segment from rod section with an average local burnup of 77 GWd/tU. During heating, the segment released approximately 6 ± 2% of generated fission gas inventory after ramp heating to 460°C. Here, heating ceased when the sample ejected from the holder, as observed by constant imaging.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Impact of grain boundary and surface diffusion on predicted fission gas bubble behavior and release in UO 2 fuel

In this work, we quantify the impact of grain boundary (GB) and surface diffusion on fission gas bubble evolution and fission gas release in UO 2 nuclear fuel using simulations with a hybrid phase field/cluster dynamics model. Here, we begin with a comprehensive literature review of uranium vacancy and xenon atom diffusivity in UO 2 through the bulk, along GBs, and along surfaces. In our model we represent fast GB and surface diffusion using a heterogeneous diffusivity that is a function of the order parameters that represent bubbles and grains. We find that the GB diffusivity directly impacts the rate of gas release via GB transport, and that the GB diffusivity is likely below 104 times the lower value from Olander and van Uffelen. We also find that the surface diffusivity impacts bubble coalescence and mobility, and that the bubble surface diffusivity is likely below 10 -4 times the value from Zhou and Olander.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

The role of irradiation-enhanced interstitial diffusion in over-pressurizing fission gas bubbles in UO 2

Fission gas bubbles in UO 2 nuclear fuel have been observed to exhibit pressures in excess of the equilibrium bubble pressure; however, the cause of bubble over-pressurization has not yet been demonstrated. The mechanical interaction between a bubble and the surrounding matrix or grain boundary depends on the internal pressure of the bubble and local stress state, such that over-pressurized bubbles are thought to be responsible for fragmentation and pulverization, when exposed to a temperature ramp. Here, in this work, we investigate the role of U interstitials, produced through irradiation, in over-pressurizing bubbles by using a combined molecular dynamics (MD) and cluster dynamics approach. Firstly, the energies for the capture of interstitials and vacancies by bubbles have been determined from MD as a function of the ratio of gas atoms to vacancies that make up the bubble. Secondly, these reaction energies have been implemented in the cluster dynamics code Centipede to predict bubble over-pressurization as a function of temperature for typical fission rates. It was found that there is a transition from low pressure bubbles (at high temperatures) to high pressure bubbles (at lower temperatures). The cause of this behavior was shown to be the creation of irradiation-induced interstitials that are highly mobile relative to vacancies at low temperature; whereas, vacancies are sufficiently mobile at high temperatures to limit bubble pressures. This result supports the hypothesis that over-pressurized bubbles form during steady-state operation and that this behavior is highly sensitive to the local pellet temperature.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Nanoscale clustering and fission product segregation in irradiated annular U-10Zr fuel

Zirconium (Zr) is added to uranium (U) to improve the performance of metallic fuel for fast reactor applications. This study employs transmission electron microscopy (TEM) and atom probe tomography (APT) to investigate nanoscale clustering of U and Zr, as well as segregation of fission products (FPs), in annular U-10Zr (in weight) metallic fuel irradiated at the Advanced Test Reactor (ATR). The results reveal variations in the shape, size, and chemical composition of clusters at different radial locations within the irradiated fuel cross-section. Zr-rich clusters exhibit higher concentration of FPs compared to U-rich clusters, potentially due to the co-precipitation of Zr and FPs in the fuel matrix during cooling at the end-of-life. In conclusion, this work complements the study of fuel constituents and fission product distribution across multiple length scales in irradiated U-10Zr metallic fuel.

Atom probe tomography↗
DOE OSTI2026

Documentation, assessment, and improvements of fission gas modeling capabilities in BISON

Fission gas behavior critically affects nuclear fuel performance, influencing thermal conductivity, microstructure, swelling, and fuel-cladding interactions. The Simple Integrated Fission Gas Release and Swelling (Sifgrs) model, implemented within the BISON fuel performance code, provides advanced capabilities for modeling these phenomena. This manuscript reviews and assesses Sifgrs’ modular structure and submodels describing intragranular, intergranular, and release mechanisms. The modular implementation enables rapid improvements as our understanding advances and new fuel forms emerge. After an overview of capabilities, we present a comprehensive validation suite, comparing predictions to experimental data from tests under normal and transient conditions, including reproduction of the Vitanza curve. Results demonstrate Sifgrs’ accuracy, robustness, and applicability. Current limitations and ongoing efforts are discussed, including mechanistic, multiscale modeling, uncertainty quantification, and validation within the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, supporting the safe and efficient use of nuclear fuels.

BISON↗
DOE OSTI2026

Evaluating transient fission gas release in high burnup light water reactor fuel during loss of coolant accident conditions via new capabilities

In this work, the role of transient fission gas release (tFGR) in the cladding burst behavior of high burnup fuel during a loss-of-coolant accident (LOCA) in commercial light water reactors was further investigated via the use of a new apparatus. During the LOCA-related temperature ramps of high burnup fuel, the release of fission gases exceeds the steady-state release observed under normal operating conditions. An enhancement was made to the Oak Ridge National Laboratory Severe Accident Test Station (SATS) to probe the various factors influencing tFGR. Experiments were performed on commercially irradiated, zirconium-clad uranium dioxide fuel, and this paper details the design of the experimental setup, the initial test results, and the subsequent post-test analyses. Notably, the first test on high burnup fuel demonstrated a LOCA-relevant tFGR of 5.3% from an unpressurized fuel segment. The ultimate tFGR was 10.7% for beyond LOCA conditions. A follow-up test on similar fuel revealed a tFGR of 12.6% under comparable conditions. Microstructural analysis and an analysis of the released gas provide some insight regarding the source of tFGR in the fuel. Finally, a grain boundary bubble model may aid in the interpretation of the results and offer a guide for future work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Nonorthogonal Configuration Interaction for Singlet Fission: Beyond the Dimer

Non-orthogonal configuration interaction with fragment calculations are presented for a number of compounds that show singlet fission properties: (i) four perylene-diimide derivatives, (ii) crystalline pentacene and its (B,N)-substituted variant, and (iii) a regular and a distorted stack of three indolonaphthyridine molecules. The electronic couplings between the singlet excitonic states (S 1 ) and the singlet-coupled double triplet (T 1 T 1 ), the so-called singlet fission coupling, were computed from ensembles with two and three molecules, and except for some small deviations when charge transfer states were included, results are virtually the same. Ensembles of three molecules were used to study the mechanisms of triplet separation, double triplet diffusion, and singlet and triplet exciton diffusion. The calculations show that apart from the standard mechanism for the generation of two uncoupled triplet states (S 1 → T 1 T 1 → T 1 ...T 1 ), there are two other possible pathways: the direct generation from the singlet excitonic state (S 1 → T 1 ...T 1 ) and the process in which the excitonic state evolves in a superposition of T 1 T 1 and T 1 ...T 1 states. Furthermore, the electronic coupling for triplet diffusion is in general much smaller than for singlet diffusion.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Singlet fission for quantum information and quantum computing: the parallel JDE model

Abstract Singlet fission is a photoconversion process that generates a doubly excited, maximally spin entangled pair state. This state has applications to quantum information and computing that are only beginning to be realized. In this article, we construct and analyze a spin-exciton hamiltonian to describe the dynamics of the two-triplet state. We find the selection rules that connect the doubly excited, spin-singlet state to the manifold of quintet states and comment on the mechanism and conditions for the transition into formally independent triplets. For adjacent dimers that are oriented and immobilized in an inert host, singlet fission can be strongly state-selective. We make predictions for electron paramagnetic resonance experiments and analyze experimental data from recent literature. Our results give conditions for which magnetic resonance pulses can drive transitions between optically polarized magnetic sublevels of the two-exciton states, making it possible to realize quantum gates at room temperature in these systems.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗