Investigation of Power Flow in a Tapered Transmission Line Modification to HERMES -III using an Array of Dose- and Dose-rate Detectors in the Near- and Far-field
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Renk, Timothy Jerome↗
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Deep Learning Approach for High-accuracy Electron Counting of Monolithic Active Pixel Sensor-type Direct Electron Detectors at Increased Electron Dose
Abstract Electron counting can be performed algorithmically for monolithic active pixel sensor direct electron detectors to eliminate readout noise and Landau noise arising from the variability in the amount of deposited energy for each electron. Errors in existing counting algorithms include mistakenly counting a multielectron strike as a single electron event, and inaccurately locating the incident position of the electron due to lateral spread of deposited energy and dark noise. Here, we report a supervised deep learning (DL) approach based on Faster region-based convolutional neural network (R-CNN) to recognize single electron events at varying electron doses and voltages. The DL approach shows high accuracy according to the near-ideal modulation transfer function (MTF) and detector quantum efficiency for sparse images. It predicts, on average, 0.47 pixel deviation from the incident positions for 200 kV electrons versus 0.59 pixel using the conventional counting method. The DL approach also shows better robustness against coincidence loss as the electron dose increases, maintaining the MTF at half Nyquist frequency above 0.83 as the electron density increases to 0.06 e−/pixel. Thus, the DL model extends the advantages of counting analysis to higher dose rates than conventional methods.
Materials Science↗
Instrumentation Design and Testing Thoughts [Slides]
Slides for a presentation discussing methodologies and instruments used to test radiation in the lunar environment.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Development of a military-specific mesh-type computational phantom library and its application to internal dosimetry and in-field radiological triage screening
Estimates of organ-absorbed and committed doses to individuals exposed to radioactive materials via acute inhalation often rely on internal dose coefficients and detector responses from reference human computational models. To achieve more accurate dose assessments to United States Armed Forces service members exposed in-field, computational models with varying morphometric parameters representative of this population are necessary. The International Commission on Radiological Protection (ICRP) Publication 145 provides detailed mesh reference computational phantoms (MRCPs) for adult males and females, with morphometric parameters matched to the 50th percentile. Previously, these phantoms were 2D and 3D scaled to match desired height, mass, and secondary anthropomorphic parameters in the creation of the University of Florida / Memorial Sloan Kettering (UF/MSK) computational phantom library. To achieve body fat percentage targets required for accession into the US Armed Forces, muscle and fat volumes were adjusted accordingly, thus, creating the UF/Department of Defence computational phantom library presented in this study. A comprehensive library of mesh-type computational human phantoms was created, including 57 adult males and 49 adult females with morphometric parameters aligned with United States Armed Forces service members. Phantoms were restricted to a body mass index between 19 and 27.5, with body fat percentages below 26% for males and 36% for females. Specific absorbed fractions were computed for selected source and target combinations, demonstrating how variations in height and body mass influence energy absorption in target regions relative to the ICRP MRCPs. Radiation detector responses were also computed, revealing that higher body masses resulted in decreased registered counts in the detection volume. These findings highlight the importance of incorporating morphometric variability in computational phantoms to achieve more accurate dose assessments and radiation detection responses for United States Armed Forces service members who inhale radioactive materials in-field.
computational phantoms↗
Progress in the design of the future circular collider FCC-ee interaction region
In this paper we discuss the latest developments for the FCC-ee interaction region layout, which represents one of the key ingredients to establish the feasibility of the FCC-ee. The collider has to achieve extremely high luminosities over a wide range of center-of-mass energies with two or four interaction points. The complex final focus hosted in the detector region has to be carefully designed, and the impact of beam losses and of any type of synchrotron radiation generated in the interaction region, including beamstrahlung, have to be evaluated in detail with simulations. We give an overview of the progress of the whole machine-detector-interface-related studies, among which are the updated mechanical model of the interaction region, the plans for a novel R&D activity of a IR mockup which is just starting, the collimation scheme and evaluation of beam induced backgrounds in the detectors, evaluation of radiation dose in the experimental area, and MDI integration with the detector.
43 PARTICLE ACCELERATORS↗
Evaluation of planar silicon pixel sensors with the RD53A readout chip for the Phase-2 Upgrade of the CMS Inner Tracker
The Large Hadron Collider at CERN will undergo an upgrade inorder to increase its luminosity to7.5 × 10$^{34}$ cm$^{-2}$s$^{-1}$. The increased luminosityduring this High-Luminosity running phase, starting around 2029,means a higher rate of proton-proton interactions, hence a largerionizing dose and particle fluence for the detectors. The currenttracking system of the CMS experiment will be fully replaced inorder to cope with the new operating conditions. Prototype planarpixel sensors for the CMS Inner Tracker with square50 μm × 50 μm and rectangular100 μm × 25 μm pixels read out by theRD53A chip were characterized in the lab and at the DESY-II testbeamfacility in order to identify designs that meet the requirements ofCMS during the High-Luminosity running phase. A spatial resolutionof approximately 3.4 μm (2 μm) is obtained using themodules with 50 μm × 50 μm(100 μm × 25 μm) pixels at the optimalangle of incidence before irradiation. After irradiation to a 1 MeVneutron equivalent fluence ofΦ$_{eq}$ = 5.3 × 10$^{15}$ cm$^{-2}$, a resolution of9.4 μm is achieved at a bias voltage of 800 V using a modulewith 50 μm × 50 μm pixel size. All modulesretain a hit efficiency in excess of 99% after irradiation tofluences up to 2.1 × 10$^{16}$ cm$^{-2}$. Further studies ofthe electrical properties of the modules, especially crosstalk, arealso presented in this paper.
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
Criticality Accident Alarm System Modeling for the Uranium Processing Facility
The Criticality Accident Alarm System (CAAS) for the Uranium Processing Facility (UPF) provides coverage for two primary processing facilities. For UPF, detailed documentation and drawings of the design are readily available. While this design information allows for some CAAS modeling features to match the actual design with a high level of accuracy, assumptions and simplifications are necessary as it is unrealistic and unnecessary to replicate many aspects of the design. A set of modeling guidelines and techniques were used to develop consistency between the MCNP6 models. While ensuring conservatism, overly conservative choices must be avoided to demonstrate the minimum accident of concern can be detected. To ensure proper modeling choices were made for the final models, a group of parametric studies were performed to evaluate the impact on the dose rates detected at the individual detectors. Some of these modeling choices and/or associated parametric studies include: - Modeling the grid steel area above the detectors as homogeneous stainless steel and determining the impact on detectability with an increasing density. - Modeling the presence of equipment in close proximity to the CAAS detectors. - Determining an appropriate material density to represent miscellaneous items not specifically modeled. - Evaluating the impact of sprinkler activation. The results of UPF CAAS analyses show that the modeling assumptions made based on the detailed design and parametric studies are appropriate and demonstrate accurate yet conservative models have been developed for detecting minimum accident of concern within the UPF.
61 RADIATION PROTECTION AND DOSIMETRY↗
Evaluation of neutron dosimetry capabilities with the MC-15 portable multiplicity counter
This work proposes a preliminary neutron dose rate estimation method for a neutron multiplicity detector through measurement- and simulation-based analyses. Uncharacterized neutron-emitting sources may be encountered in situations such as nuclear emergency response, safeguards, and treaty verification. These circumstances may present irradiation risk to personnel conducting field assay, search, and characterization measurements. It is therefore of interest to provide a field neutron dosimetry capability with the existing neutron multiplicity counting (NMC) capabilities. To date, no commercially-available neutron detection systems are capable of both accurate NMC and real-time neutron dosimetry. This work will focus on estimating dose rate using input from a single fielded NMC called the MC-15. The energy-dependent neutron detection efficiency response of the MC-15 was quantified in monoenergetic neutron simulations and evaluated in response to two neutron-emitting sources and to a polyethylene-moderated source. The results were compared to existing neutron dosimeters and established the proof of concept for further investigation of the MC-15 for dose estimation. Measurement results were also replicated in simulations; additional simulations were then conducted to expand upon the limited empirical data. The initial empirical results provided a conversion factor appropriate for use when measuring 252 Cf neutrons that is independent of polyethylene shielding presence and thickness. The simulated data sets were then used to evaluate a fit equation allowing estimation of the neutron dose rate for less restricted geometric configurations and dependent only on the distance between the source and the detector. Additionally, the energy dependence of the efficiency response indicates that further empirical evaluations could provide energy-dependent conversion factors for broader neutron dosimetry capabilities with a wider range of neutron-emitting sources.
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
Effects of Proton Irradiation on the Performance of Skipper CCDs
Skipper CCDs are a mature detector technology that has been suggested for future space telescope instruments requiring sub-electron readout noise in the near-ultraviolet to the near-infrared. While modern skipper CCDs inherit from the radiation-tolerant p-channel detectors developed by LBNL, the effects of high doses of ionizing radiation on skipper CCDs (such as those expected in space) remains largely unmeasured. We report preliminary results on the performance of p-channel skipper CCDs following irradiation with 217-MeV protons at the Northwestern Medicine Proton Center. The total nonionizing energy loss (NIEL) experienced by the detectors exceeds 6 years at the Sun-Earth Lagrange Point 2 (L2). We demonstrate that the skipper amplifier continues to function as expected following this irradiation. Owing to the low readout noise of these detectors, controlled irradiation tests can be used to sensitively characterize the charge transfer inefficiency, dark current, and the density and time constants of charge traps as a function of proton fluence. We conclude with a brief outlook toward future tests of these detectors at other proton and gamma-ray facilities.
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
Evidence for ionization damage in mid-wave infrared nBn detectors
The effects of 63 MeV proton and 60Co gamma irradiation on the operation of 4.3 μm cutoff nBn photodetectors are demonstrated separately, and both are shown to yield a total ionizing dose (TID) effect. The effect is shown here in an InAsSbBi nBn detector and has been observed in other bulk alloy nBn detectors, and is unusual as it is notably absent in superlattice nBn detectors. The non-antireflection coated detectors exhibit a pre-radiation quantum efficiency of 17% at 3.3 μm wavelength and a dark current density of 50 μA/cm2, or roughly 300× the Rule 07 expectation, at their ideal operating voltage of −0.4 V bias at 150 K. Step-wise proton irradiation and in situ measurement indicate that the dark current increases to about 400× Rule 07 at the highest proton dose level of 150 krad(Si) (9.10 × 1011 p+/cm2), while the quantum efficiency is degraded at a relatively faster rate than the majority of analogous detectors characterized by our lab. Both the photocurrent and dark current are also shown to exhibit a turn-on voltage magnitude reduction of 100 mV following either gamma or proton irradiation, a trend which is attributable to negative trapped charge at the barrier interface (TID effect). This theory is further supported by an observed capacitance density magnitude reduction with dose and affirmed with Silvaco TCAD simulations. Following both proton exposure and subsequent anneal and gamma exposure and subsequent anneal; dark current, photocurrent, and CV all approach their pre-radiation baseline values.
Physics↗
SCALE 6.2.4 Validation: Radiation Shielding
For safe and reliable use of computer codes by the community, accuracy must be clearly evaluated. In particular, the nuclear reactor engineering and licensing field needs accurate tools for radiation shielding modeling. Monaco with Automated Variance Reduction using Importance Calculations (MAVRIC) is one such tool, with built-in variance reduction methods distributed within the SCALE code, and its validity is demonstrated in this report for the released version 6.2.4. Representative benchmarks corresponding to shielding analysis are selected for the validation study. Typical experimental results analyzed from those benchmarks include neutron fluxes, detector count rates, detector energy response functions, neutron and gamma doses, foil neutron activation rates and activities, neutron leakage fluxes, and skyshine dose rates. Thousands of points of comparison between experiment and calculation are presented in this work. Other than rare outliers typically explained by either a lack of information or large uncertainties in the experiment conditions, material, or dimensions, MAVRIC agrees well with the experiment results. MAVRIC is also compared to Monte Carlo N-Particle (MCNP) calculations when available, and both codes generally produce good agreements within estimated uncertainties. The selected benchmarks are obtained from reliable sources such as the International Criticality Safety Benchmark Evaluation Project Handbook (ICSBEP Handbook), the Shielding Integral Benchmark Archive & Database (SINBAD), and other shielding validation work found in the literature. Additional datapoints and benchmarks will be added to future versions of this report to incrementally expand the shielding validation suite incrementally.
61 RADIATION PROTECTION AND DOSIMETRY↗
SCALE 6.3 Validation: Radiation Shielding
Safe and reliable use of scientific and engineering computer codes requires validation for the types of applications in which they will be used. An example in the nuclear reactor engineering and licensing field is radiation transport employed in shielding analyses. The validity of computer codes for shielding applications is demonstrated in this report for SCALE version 6.3.0. Representative benchmarks corresponding to shielding analyses are selected for the validation study. Typical measurement results analyzed from these benchmarks include neutron fluxes, detector count rates, detector energy response functions, neutron and gamma dose rates, neutron activation rates and activities, neutron leakage fluxes, and skyshine dose rates. Thousands of points of comparison between measurement and calculation are presented in this work. Other than rare outliers typically explained by either a lack of information or large uncertainties in the experiment conditions, material, or dimensions, the Monaco with Automated Variance Reduction using Importance Calculations (MAVRIC) radiation transport computer code with built-in variance reduction methods distributed with the SCALE computer code system agrees well with the measurement results. In selected benchmarks, MAVRIC is also compared to Monte Carlo N- Particle® (MCNP® ) 1 calculations. Both computer codes generally agree well within the estimated uncertainties. With the release of SCALE 6.3.0, Shift was integrated as an alternative transport solver in MAVRIC, denoted MAVRIC-Shift. Although the traditional MAVRIC using Monaco was used primarily in this validation study, many results have also been generated using MAVRIC-Shift. Agreement between MAVRIC-Monaco and MAVRIC-Shift is generally very good. The benchmarks presented in this report were obtained from reliable sources such as the International Criticality Safety Benchmark Evaluation Project Handbook, the Shielding Integral Benchmark Archive & Database, and other shielding validation work found in the literature. Additional datapoints and benchmarks will be added to future versions of this report to expand the shielding validation suite.
61 RADIATION PROTECTION AND DOSIMETRY↗
Griffin Capability Improvements in Support of Ex-core Deep-Penetration Problems
Advanced reactor designs, especially portable reactors that are designed to be located closer to humans and operate autonomously, require the ability to accurately compute the ex-core neutron and gamma flux solutions in terms of shielding design optimization to reduce dose rates at the vessel boundary and detector signal prediction to drive the reactor control system. The Nuclear Energy Advanced Modeling and Simulation program has prioritized improvements to the Griffin discrete ordinates (SN) solver for deep-penetration problems in fiscal year 2025. Significant advancements have been made to the Griffin methodologies for solving ex-core deep-penetration problems for steady-state, fixed-source and transient calculations. This work presents the methodology improvements as well as a comprehensive demonstration with a Transient Test Reactor model and measurements.
22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
Gamma and Neutron Measurement and Modeling of Irradiated TRISO Fuel
Given the unique characteristics of the PBR fuel cycle, both gamma and neutron measurements are expected to play important roles in performing and maintaining nuclear material control and accounting for spent pebbles to safeguard the fuel cycle. Given the lack of irradiated pebbles in the US, a variety of irradiated TRISO fuel samples with wide ranges of burnups and cooling times available at ORNL were used in this work. A large number of gamma and neutron measurements have been performed on these samples to collect data to test the various detectors and to benchmark the computer models to simulate the depletion and decay of the fuel and the measurements themselves. Two neutron detectors, including a custom-made detector and the Very High-Performance Neutron Multiplicity Counting, were used to measure the neutrons emitted by these TRISO samples. Three gamma spectrometry detectors, including an HPGe and the M400 CZT detector, were used to measure gamma-ray emissions from these samples. The M400 was recently adopted by the IAEA for fresh uranium measurements, but it was tested for spent fuel measurements prior to this project. Detailed MCNP models were developed to simulate these neutron and gamma measurements. Some GADRAS models were also developed to cross check the MCNP models for the gamma measurements. It was found challenging to perform neutron measurements in the hot cell due to the high background counts. Close agreements were observed between the simulated and measured neutron count rates in both detectors’ measurements of californium calibration sources. Both the HPGe and M400 detectors were able to measure the 604 and 662 keV peaks from these samples, which are the two most important peaks used to infer fuel burnup. Although the M400 detector did not have nearly good energy resolution and did not detect some of the minor peaks as the HPGe detector, it was found to be capable of handling significantly higher dose rates than HPGe. Given the complexities in the TRISO samples (e.g., different samples sizes) and uncertainties in the alignments between the detector and the TRISO fuel inside the containers, large scatters were found between the peak area rates and the samples’ burnups. However, the 604/662 peak ratios were found to trend well with the samples’ burnups among most samples in both measured and simulated results.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Arbitrary electron dose waveforms for electron microscopy
A device may include an electron source, a detector, and a deflector. The electron source may be directed toward a sample area. The detector may receive an electron signal or an electron-induced signal. A deflector may be positioned between the electron source and the sample. The deflector may modulate an intensity of the electron source directed to the sample area according to an electron dose waveform having a continuously variable temporal profile.
Bloom, Ruth Shewmon↗
Effects of 63 MeV proton irradiation on the performance of MWIR InGaAs/InAsSb nBn photodetectors
An investigation into the effects of 63 MeV proton irradiation on high-sensitivity mid-wave infrared InGaAs/InAsSb nBn devices is performed. Three different structures with various absorber region doping profiles are irradiated and characterized to assess their impact on performance degradation. Minority carrier lifetime is measured using time-resolved photoluminescence and lifetime damage factors are assessed. The majority carrier concentration is determined via capacitance–voltage measurements and dopant introduction rates are calculated. An analysis of dark current density is performed using these material parameters, revealing a reduction in mobility with proton fluence and the emergence of a proton-induced trap energy level. Quantum efficiency is calculated at each proton fluence, and quantum efficiency damage factors show that the graded doping structure exhibits the least reduction of quantum efficiency with dose, attributed to its effective mobility enhancement. Conclusively, detector sensitivity, assessed via shot-noise limited noise-equivalent irradiance, shows that the graded doping structure is the least susceptible to high energy proton irradiation-induced performance degradation.
Physics↗
Energy dependent spectrum-to-exposure conversion coefficients for Mobile and Gemini systems
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61 RADIATION PROTECTION AND DOSIMETRY↗
SFRC Measurement Control Summary - FY23 Q3/Q4
Measurements with the Spent Fuel Rod Counter (SFRC) to be performed bi annually include 3 He detector efficiency checks (in both modules) and ion chamber channel tests with a pocket pulser. These measurements will all be performed at TA-35, building 27 where the SFRC is stored. Refer to LA-UR-22-31737 and LA-CP-22-20079 SFRC measurement control documents for a description of additional tests. Note that during this measurement period, counts could not be confirmed for the Si detectors, as an analog pulse to TTL box necessary for the DAQs to communicate with the detectors was still under construction. For this reason, typical measurements at the Calibrated Dose Facility were not performed for the second half of FY23. When these components are finished by the NEN electrical shop, the Si detectors will be implemented into these measurement control checks.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗