Characterization of Neutron Sources using Foil Activation
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Measured Reaction Rate Ratios of 235 U, 238 U, 237 Np, 239 Pu Samples in the PFUNS 235 U Prompt Fission Neutron Spectrum Criticality Experiment
The Prompt Fission Uranium Neutron Spectrum experiment, an experiment to reduce uncertainties in the high energy tail of the 235 U prompt fission spectrum, achieved success by performing two separate irradiations measuring approximately 40 different IRDFF-II reactions total using over 20 different foil materials at the National Criticality Experiments Research Center in February 2024. The criticality experiment utilized a set of highly enriched uranium hemispherical shells of increasing diameters with a large void in the center where the samples were located. The focus of this work is the first of two PFUNS irradiations focused on irradiating two of each fission foils, one bare and one cadmium covered, along with metal activation foils containing reaction products with short half-lives, such as indium, iron, and gold along with nickel fluence monitors. This work focuses on presenting the initial reaction rate ratio results of the fission foils from the aforementioned first irradiation to assist in nuclear data validation of those species in a nearly pure 235 U prompt fission neutron spectrum and compares to the Lady Godiva and Flattop-25 historic experiments at the Los Alamos Critical Experiments Facility. Future work will combine fission foil and metallic activation foil reaction rate ratio results from both the first and second higher power irradiation and reaction rates for a final spectral adjustment.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
AGC-4 As-Run Uncertainty Analysis
The concept of measuring neutron flux (or fluence rate) with activation foils has existed since the late 1940s and is a well-used methodology. Activation foils are selected for their unique isotopic cross sections that create reaction products with desirable properties: relatively simple radioactive decay scheme, measurable quantities of gamma-rays or betas, and half-lives appropriate for the irradiation and decay durations of the experiment. The activation materials, deemed flux wires, fluence wires, dosimeters, or flux foils, are fabricated to have a mass appropriate for the irradiation as well so the measurement using a radiation detector following the irradiation is sufficient but not overwhelming in terms of total activity. Activation measurement results are expressed in terms of reaction rate, which can then be used to “adjust” the neutron spectrum. The Advanced Graphite Creep (AGC)-4 experiment contained 26 flux wire packages, each containing an iron and a niobium wire inside a vanadium capsule. For all stackup positions, the value of fast to thermal neutrons is higher than expected, but the total neutron fluence is near the predicted value of ~2 x 1022 n/cm2. As compared to the predicted neutron spectrum, there are four takeaways in the results as compared to the predicted ATR irradiation: 1) Additional fast (fission) neutrons present, 2) Spectrum hardening due to temperature, 3) Spectrum hardening due to container attenuation, 4) Total neutron fluence is close to predicted values. Since the neutron energy spectrum is a key parameter when attributing neutron damage in materials, the adjusted energy-dependent spectrum and the total neutron fluence results from the dosimeters can be incorporated for future calculations.
11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
LaBr 3 : Ce self-activation analysis for measuring fast neutron fields
Measurement of the fast neutron production rate in deuterium–tritium (D–T) fusion reactions is important for applications such as active interrogation, fusion diagnostics, and borehole logging. Such measurements are typically performed by neutron activation analysis of metal foils, especially copper. Copper foil activation analysis requires efficiency and energy calibrations of the detector used to measure the foil, and it relies on the detection of 511 keV gamma rays, which are prominent in the active background when neutrons are being produced. Alternatives, such as 79m Br produced by inelastic neutron scattering on 79 Br, are limited by short half-life, low-energy gamma emission, and inability to selectively measure D–T neutrons. This work describes a novel alternative approach to measure ≳10 MeV neutron fields based on self-activation analysis of a LaBr 3 :Ce detector. The activity of 78 Br, the activation product of the 79 Br(n,2n) 78 Br reaction, is used to determine the neutron flux and infer the neutron production rate. We experimentally demonstrate the method with a cylindrical LaBr 3 :Ce crystal with a diameter and height of 3.81 cm that was placed at an ~18 cm distance from the neutron production point, at a 90° angle with respect to the deuteron beam in a D–T neutron generator. Operating voltage and current of the generator were adjusted to evaluate the technique’s performance over a nominal generator output range of approximately (1 - 9) x 10 7 n/s. The neutron output obtained from LaBr 3 :Ce activation agrees to within three standard deviations of the output obtained using copper activation. The self-activation technique can be conveniently employed in a variety of applications to simplify measurements of fast neutrons produced in D–T fusion reactions.
Active interrogationLaBr3↗
Godiva IV Simulated Radiation Field Characterization and Variance Reduction
Godiva IV is a system comprised of highly enriched uranium alloyed with molybdenum in the form of fuel plate rings. The reactor, along with its predecessors, was designed with the unique ability to satisfy interests in the super-prompt-critical reactor operation space. Originally, the reactor was part of the Los Alamos Critical Experiments Facility (LACEF) at Technical Area-18 (TA-18). The radiation field around Godiva at this facility was well characterized and understood. As a fast neutron system, the neutron spectrum in and around Godiva was close to a Watt Fission spectrum. The Kiva where Godiva IV was located at LACEF was made of thin, sheet metal walls which did not contribute significantly to the neutron spectrum. Following the transition of LACEF to the National Critical Experiments and Research Center (NCERC) in Nevada, Godiva-IV was moved from TA-18 to the Device Assembly Facility (DAF) at the Nevada National Security Site (NNSS). Part of this move brought renewed interest in radiation field characterization. The new facility introduced significant changes to the environment surrounding Godiva, and preliminary foil irradiation results suggested that the room contribution to the neutron spectrum was significant. Unlike at TA-18, a large thermal neutron signature was added to the fast spectrum from Godiva due to significant room return. A primary goal due to the additional complexity that the room return adds to the Godiva IV radiation emission spectrum was the development of an efficient Monte Carlo N-Particle (MCNP) calculation capable of characterizing the neutron spectrum anywhere in the room around Godiva. A campaign of activation foil irradiations and analysis were completed to support the validation of the MCNP model. The modeling of these foils in MCNP can be easily done with a standard volumetric neutron flux tally. However, given the multitude of locations and reaction rates to be modeled, further steps must be taken to increase the efficiency of these calculations in MCNP. During this study, a benchmark model currently under development for Godiva IV was used. A qualitative assessment of the thermal neutron contributors was performed using spatial neutron distribution plots. Additional detail was added to the model based on the qualitative results showing the thermal spectrum’s large sensitivity to hydrogenous material. Neutron energy spectra was evaluated at discrete locations in the room around Godiva to quantify the relative contribution of various components. It was discovered that the concrete walls are the largest contributor to the thermal signature, with minor contributions from plastic components surrounding Godiva. Following these results, two different variance reduction techniques were implemented to improve the problem efficiency in these calculations. In the first approach, an F5 point detector tally was implemented in the standard Godiva IV criticality problem. The second approach involved a weight-window generator implementation with an F5 point detector tally in a fixed source problem. The weight window implementation reduced the runtime from 42739.55 minutes to 1803.34 minutes (computer time), compared to the F5 KCODE implementation.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Characterizing secondary neutrons at BLIP for isotope production applications
Fast secondary neutrons created at the Brookhaven Linac Isotope Producer (BLIP) facility following proton irradiation were characterized by the foil activation technique and compared with FLUKA Monte Carlo simulations. The FLUKA-simulated neutron flux was spectrally adjusted following the maximum entropy formalism using the International Reactor Dosimetry and Fusion File (IRDFF-II), with predictions agreeing with experimental measurements to within 9 % following the adjustment procedure. Here, a multitude of degrader configurations were simulated to assess the feasibility of improving the fast (E n > 20 MeV) secondary neutron yield at the proposed neutron target position (“N-slot”). A configuration where the N-slot is closest to the proton degrader produced the highest fast neutron yield, with tungsten degraders achieving the best performance. Assuming the optimized target-degrader configuration proposed in this work, we discuss potential isotope production opportunities with secondary neutrons. In most cases the yields are in the order of several mCi.
07 ISOTOPE AND RADIATION SOURCES↗
59 Co(p,X) spallation reaction cross sections for 250 MeV to 2 GeV protons
Cobalt is an advantageous target for probing the physics of nuclear spallation, because it is a naturally mono-isotopic element ( 59 Co), and its per-nucleon binding energy (BE/A = 8.768 MeV) is near the maximum value for all nuclei. We measured nuclear spallation cross sections for the 59 Co(p,X) reaction at five kinetic energies ranging from 250 MeV to 2 GeV. Cross sections for the production of 58 Co, 57 Co, 57 Mn, 56 Co, 56 Mn, 56 Cr, 55 Fe, 53 Fe, 54 Mn, 52 Mn, 51 Cr, 49 Cr, 48 V, 47 Sc, 46 Sc, 44 Sc, and 44 Scm are reported. Where comparable data exist in the EXFOR reaction database, we find that our measured cross sections generally agree. In many cases, we provide data for reactions or energies not currently reported in EXFOR. Our cross sections also provide evidence for the presence of α-clusters within the 59 Co nucleus, a surprising result given the asymmetry in Z (27) and N (32) for this nucleus. Finally, we use our measurements to evaluate the accuracy of spallation cross section simulations from GEANT4-based radiation transport toolkit, performed with the INCLXX-, Bertini-, and Binary-ion-cascade (BIC) based physics lists. This benchmarking activity revealed that the simulations overestimated the cross sections by a factor of ∼2–4 on average, and that the INCL-XX physics list provides the most reliable results. This evaluation informs the selection of the GEANT4 physics lists used for the analysis of data from NASA’s Psyche mission, which will measure γ rays and neutrons resulting from spallation reactions occurring on the surface of an asteroid whose surface is thought to be rich in iron-nickel metal.
43 PARTICLE ACCELERATORS↗
Conceptual Design of a Thermal–to-14-MeV Neutron Conversion Device for Use in the Advanced Test Reactor
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.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Neutron production and dosimetry with D(D,n) 3 He, 12 C(D,n) 13 N and 7 Li(p,n) 7 Be reactions at the Sandia Ion Beam Laboratory
Neutrons are produced at the Sandia Ion Beam Laboratory (IBL) using a beam of 280 keV D 2 + ions onto an erbium deuteride target. At test locations in the forward direction these neutrons have energies of about 2.9 MeV and a maximum flux on the flange of the test chamber of about 10 10 n/cm 2 /hour. Dosimetry is done by counting associated charged particles and by nickel foil activation dosimetry. Neutrons with higher energies, up to 6 MeV are produced using higher energy deuterium, up to 3 MeV, from the Pelletron ion accelerator at the IBL on a deuteride target. This report provides detailed information on the methods used for the production and dosimetry of neutrons by the D(D,n) 3 He nuclear reaction at the IBL. Neutron production by the 7 Li(p,n) 7 Be reaction with a proton beam from the Pelletron on a lithium target, such as LiF is also discussed. This produces neutrons with lower energy, in the range from about 0.1 to 1.3 MeV. Neutron production by the 12 C(D,n) 13 N reaction is also discussed. The purpose of this report is to provide information needed to test the effects of neutron irradiation on semiconductor materials and devices at the Sandia IBL.
36 MATERIALS SCIENCE↗
Preliminary Activation Wire Analysis for the POKER Experiment Flattop Irradiation
Activation foils, or in this case wires, are commonly deployed to critical experiments at the Na tional Criticality Experiments Research Center (NCERC) to provide additional neutron diagnostics and provide additional energy-dependent responses to tie back to neutronic and/or multi-physics simulations.
07 ISOTOPE AND RADIATION SOURCES↗
First experimental determination of the ⁴⁰Ar(𝑛, 2𝑛)³⁹Ar reaction cross section and ³⁹Ar production in Earth’s atmosphere
The cosmogenic ³⁹Ar(t1∕2 = 268 years) isotope of argon is used for geophysical dating and tracing of underground and ocean water, as well as ice owing to its appropriate half-life and chemical inertness as a noble gas; ³⁹Ar serves also in nuclear weapon test monitoring. We measured for the first time the total cross section of the main ³⁹Ar cosmogenic production reaction in the atmosphere, namely ⁴⁰Ar(𝑛, 2𝑛)39 Ar, using 14.8 ± 0.3 MeV neutrons. The neutrons, produced by a deuterium-tritium generator, impinged on a stainless steel sphere filled with Ar gas highly enriched in the ⁴⁰Ar isotope and were monitored by a stack of fast-neutron activation foils. The reaction yield was measured by atom counting of long-lived ³⁹Ar with noble gas accelerator mass spectrometry and, independently, by decay counting relative to atmospheric argon (³⁹Ar/Ar= 8.12 × 10−16 ). A total ⁴⁰Ar(𝑛, 2𝑛)39 Ar cross section of 610 ± 100 mb was determined at 14.8 ± 0.3 MeV incident neutron energy. This result serves as a benchmark for recent theoretical calculations and evaluations, found to reproduce well the experimental total cross section. We use these energy-dependent theoretical cross sections together with experimental spectra of cosmogenic neutrons at different altitudes to calculate the global average rate of neutron-induced ³⁹Ar atmospheric production, resulting in 770 ± 240 39 Ar atoms/cm²/day. The secular equilibrium between the ³⁹Ar calculated production rate and radioactive decay rate leads to a partial isotopic abundance ³⁹Ar/Ar = (5.9 ± 1.8) × 10−16 , showing that ≈73% of atmospheric ³⁹Ar is produced by cosmogenic neutrons, the remaining part believed to be induced by muons and high-energy 𝛾 rays. The ⁴⁰Ar(𝑛, 2𝑛)³⁹Ar cross section at 14 MeV is also a key parameter for quantifying the anthropogenic contribution to atmospheric 39 Ar produced during the thermonuclear tests of the 1960s. We estimate that anthropogenic ³⁹Ar accounts for roughly 20% of the present atmospheric inventory.
39 Ar atmospheric production↗
NCERC-CL TREAT Free Field Characterization Foil Reaction Rate Results
This report summarizes the analysis and preliminary MCNP reaction rate modeling of free field characterization measurements performed in the Transient Reaction Test Facility (TREAT) at Idaho National Laboratory (INL) between 12/18/2023 and 2/22/2024. A series of seven irradiations were performed using the Big-BUSTER (Broad Use Specimen Transient Experiment Rig) core configuration. Foil sets including 19.5% LEU-Zr alloy wire, S, Au, Fe, Ni, Co, Ti, and Zr were deployed in each irradiation. Identical foil sets were supplied to Los Alamos National Laboratory (LANL), Lawrence Livermore National Laboratory (LLNL), and INL. All foils were supplied by INL. Additional details on the experiment can be found in [2]. The foil sets were shipped to the National Criticality Experiments Research Center Counting Laboratory (NCERC-CL) branch at LANL (NCERC-CL NISC) and were received on 3/25/2024. Additional details on receipt and counting are reported in [3]. All reactor dosimetry measurements were performed in adherence to the ASTM standards applicable to reactor dosimetry. Measured reaction rates from Au, Co, Fe, Ni, and Ti are reported. Select reaction rate ratios and simulated reaction rates and ratios are discussed. Reaction rates for all measured reaction products were modeled with MCNP 6.3.1 and a TREAT input deck supplied by Edward Lum. Neutron emission estimates calculated using the measured and simulated reaction rates show excellent agreement between the three fast-threshold n-p reactions with the percent differences being < 9% between the three reactions. The capture reactions had poorer, but still reasonable, agreement with < 20% percent differences between the neutron emission estimates of the three capture reactions. The percent difference between the fast-threshold and capture reactions is ∼ 100% for each reaction. This supports the TREAT model is not accurately modeling the neutron spectrum and additional measurements are required to characterize TREAT and match measurements to simulation.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Au-198 and TLD-600 results from IER-593 Godiva irradiations
This report summarizes the results from dosimetry packets deployed during a nominally 250 ◦C temperature rise Godiva IV burst. Dosimetry packets including Bare Au foils, Cd covered Au foils, and Thermoluminescent Dosimeters (TLDs) were deployed at five positions around Godiva at a height corresponding to the center height of the HEU fuel.
61 RADIATION PROTECTION AND DOSIMETRY↗
Experimental observation and integrated modelling of proton-beryllium fusion in He and D plasmas at JET
Validated integrated modelling of JET ITER-like wall experiments in which fusion performance is driven by reactions between fast ions and intrinsically present metal wall impurities is presented. A steady-state L-mode plasma with dominant proton-beryllium fusion and neutron yields of up to ≈ 6·10 13 s -1 is developed in He and D, via radiofrequency heating of a H minority. The fusion drive is unambiguously confirmed by the neutral particle analyser, fast ion loss detector, and γ-ray diagnostics. Experiments are analysed via an integrated modelling framework, developed to model the two-stage proton beryllium-fusion chain and produce high-fidelity fusion product source terms. The modelling chain comprises TRANSP and JETTO for plasma core modelling, LOCUST for full orbit product tracking and collisional slowing-down, DRESS to resolve two- and three-body fusion kinematics, and MCNP for neutron transport calculations. Modelling shows that the primary 9 Be(p,n) 9 B reaction is the dominant neutron emitter at naturally present concentrations of beryllium in these experiments. The yield contribution of secondary reactions between fusion products and beryllium, 9 Be(d,n) 10 B and 9 Be(α,n) 12 C, is found to be negligible. The proton-deuteron knock-on effect in D plasmas is modelled, which is calculated to contribute ≈ 25% to the total neutron yield. For both He and D discharges the total computed neutron rates match fission chamber (FC) measurements within the combined experimental and computational uncertainty, with an average discrepancy of ≈ ± 20%. Realistic proton-beryllium neutron sources are propagated through JET’s MCNP neutron transport model which shows that 235 U FCs’ response is sensitive to p–Be source changes, with up to ≈ 10% variation compared to a D–D neutron source. We show that the high-energy tail of the fast proton minority can be studied with multi-foil neutron activation. The framework is also applied to the study of interactions between fast protons and boron impurities, of relevance to ITER. We calculate that in JET conditions a significant alpha source with DT-like energies could be generated through 11 B(p, α)2α fusion, and detected via γ-emission in secondary interactions between fast alphas and boron. The work represents an important step towards validating predictive integrated modelling capabilities for non-standard fusion reactions.
JET↗
FLUKA simulation validation to measure the beam energy of a cyclotron
FLUKA is a fully integrated particle physics MonteCarlo simulation package which has many applications in high energy experimental physics and engineering. When using a cyclotron for medical radionuclide production, it is important to know that the beam energy is suitable to cause the intended nuclear reaction and minimize unwanted impurities. Here, this work focuses on the comparison of FLUKA as a tool to measure the beam energy of a cyclotron with experimental results using the method of stacked copper foils interspaced by an aluminum degrader. We compared two methods to determine the activity ratio: determining the activity in each foil using the experimentally measured cross-sections from the National Nuclear Data Center (NNDC) cross-section data and calculating the activity in the two Cu foils using the FLUKA program. These results were compared to those reported by Gagnon et al., 2011. The FLUKA results gave energy measurements that were different than the true energy by a percentage in the range of 2.4% to 13%, suggesting that FLUKA is not recommended for pre-calculating the cyclotron beam energy using these low Z materials without experimental confirmation. It also confirms that the experimental cross-section data from the NNDC database is superior for this application giving energy values that ranged from 0.1% to 1.6% from the true energy.
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
A Data-driven approach to Core Power distribution reconstruction in a Nuclear Reactor
This report presents the initial development of a data-driven approach for reconstructing the core power distribution in a nuclear reactor (power shape synthesis) using ex-core sensors. Traditional techniques rely on deploying a large number of detectors throughout the reactor core. However, this approach is not feasible for innovative reactor concepts like Advanced Reactors and Microreactors. First, the tight lattice pitch, designed to maximize power density, limits the space available for sensors. Secondly, the harsh operating conditions are not compatible with commercially available detectors. The method proposed in this work integrates high-fidelity modeling with data-driven techniques to accurately reconstruct power distribution across various reactor types, thereby reducing the reliance on in-core sensors. Purdue University Reactor One (PUR-1) was selected as the test case. The CAD model representing the latest configuration of the PUR-1 core was imported into the OpenMC simulation framework, and the model was built. Additionally, the previously developed MCNP6 model was updated. The two models were assessed against the data collected during an experimental campaign conducted in July 2024. Thirty gold foils were placed in three Irradiation Assemblies in PUR-1 core. Using the measured activity of the irradiated foils, the neutron flux at different core locations was estimated.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Preliminary analysis of TREAT free-field experiments using OpenMC
This work analyses activation calculations for dosimetry materials during a steady-state irradiation in the Transient Reactor Test (TREAT) reactor core. Hence, we developed a workflow based on the Monte Carlo code OpenMC alongside a custom depletion solver. The irradiation-induced activity as a function of time is computed, and several sensitivity studies are performed to evaluate uncertainty. This study has shown activity computations are sensitive to flux amplitude, irradiation time, atoms quantity and microscopic cross sections. Stochastic uncertainties have been propagated to evaluate the activity uncertainty for each dosimetry material. Most uncertainties are below our target of 3%, which demonstrates OpenMC as a powerful predictive and analysis tool. The precise results obtained through this newly developed computation scheme will be used in future experiments to characterize quantities of interest when operating the TREAT reactor in new configurations.
61 - RADIATION PROTECTION AND DOSIMETRY↗
Platinum and Gold Supported on Transition Metal Nitrides for Hydrogen Evolution in an Alkaline Electrolyte
Here, as the urgency to reduce reliance on fossil fuels increases due to carbon dioxide emissions, hydrogen produced by renewably powered water electrolysis has emerged as a promising technology. Alkaline electrolyzers typically exhibit lower current densities than acidic electrolyzers due to the slow kinetics of the hydrogen evolution reaction (HER) under alkaline conditions. This work developed Pt- and Au-modified transition metal nitride (TMN) thin films for improving alkaline HER kinetics. One monolayer Pt–VN, Pt–Mo 2 N, and Pt–TiN were the most promising thin-film catalysts, with alkaline HER activity approaching that of a bulk Pt foil. Additionally, the Gibbs free energy of adsorbed hydrogen was identified as a useful descriptor for alkaline HER activity on TMN and TMN-supported catalysts and has the potential to guide future studies on TMN-based catalysts for enhancing alkaline HER. For practical applications, the thin-film catalysts were then extended to Pt- and Au-modified TMN powders for alkaline HER. Both 5 wt % Pt/TiN and 2 wt % Pt/TiN powders exhibited lower overpotentials at 5 mA/cm 2 when normalized by the Pt electrochemical surface area than the commercial 5 wt % Pt/C benchmark, suggesting a Pt–TiN synergy that creates opportunities for more cost-effective alkaline HER cathodes. Moreover, 20 wt % Au/Mo 2 N also displayed an enhancement in HER activity when compared to the commercial 20 wt % Au/C benchmark.
58 GEOSCIENCES↗