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All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

HYSPLIT/MACCS Atmospheric Dispersion Model Technical Documentation and Benchmark Analysis

The nuclear accident consequence analysis code MACCS has traditionally modeled dispersion during downwind transport using a Gaussian plume segment model. MACCS is designed to estimate consequence measures such as air concentrations and ground depositions, radiological doses, and health and economic impacts on a statistical basis over the course of a year to produce annualaveraged output measures. The objective of this work is to supplement the Gaussian atmospheric transport and diffusion (ATD) model currently in MACCS with a new option using the HYSPLIT model. HYSPLIT/MACCS coupling has been implemented, with HYSPLIT as an alternative ATD option. The subsequent calculations in MACCS use the HYSPLIT-generated air concentration, and ground deposition values to calculate the same range of output quantities (dose, health effects, risks, etc.) that can be generated when using the MACCS Gaussian ATD model. Based on the results from the verification test cases, the implementation of the HYSPLIT/MACCS coupling is confirmed. This report contains technical details of the HYSPLIT/MACCS coupling and presents a benchmark analysis using the HYSPLIT/MACCS coupling system. The benchmark analysis, which involves running specific scenarios and sensitivity studies designed to examine how the results generated by the traditional MACCS Gaussian plume segment model compare to the new, higher fidelity HYSPLIT/MACCS modeling option, demonstrates the modeling results that can be obtained by using this new option. The comparisons provided herein can also help decision-makers evaluate the potential benefit of using results based on higher fidelity modeling with the additional computational burden needed to perform the calculations. Three sensitivity studies to investigate the potential impact of alternative modeling options, regarding 1) input meteorological data set, 2) method to estimate stability class, and 3) plume dispersion model for larger distances, on consequence results were also performed. The results of these analyses are provided and discussed in this report.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2022

Neutron Leakage Spectra Sensitivities for ICSBEP Benchmarks

Neutron leakage spectra have been measured, simulated, and investigated by many groups. These spectra have many uses, including determining shielding requirements and calculating dose for radiation protection purposes, validating nuclear data, and determining material composition. It has even been proposed to use measurements of neutron leakage spectra to determine the soil composition of Mars. As part of the Los Alamos National Laboratory (LANL) Experiments Underpinned by Computational Learning for Improvements in nuclear Data (EUCLID) Laboratory Directed Research Development (LDRD) project, the authors are working on developing methods to calculate sensitivities to neutron leakage spectra. This may allow nuclear data evaluators to better use neutron leakage spectra data to constrain or adjust nuclear data. This may also be particularly useful as many neutron leakage spectra measurements do not require fissile material and can be performed with well characterized neutron sources. Neutron leakage spectra measurements are usually performed by placing a detector at the outside of a nuclear system. This may be a reactor, a neutron generator, or a neutron source. Certain detection systems can use pulse height data to infer neutron energies, other detector systems rely on other ways of determining neutron energy (for example, a Bonner sphere with multiple moderator thicknesses can be used to measure neutron spectrum). These measurements typically rely on some unfolding of the measured results. For pulsed systems like the "Livermore Pulsed Spheres," time-of-flight information can also help to constrain the neutron energy spectra data.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Validation of Spent Fuel Dose Rate Calculations Using NDA Measurement Campaign Data

The predicted radiation gamma dose rate is a key parameter to determine if a spent nuclear fuel assembly is self-protected from misuse as well as an indicator for worker/inspector safety. Recent studies showed that using state-of-the-art techniques for dose rate calculations, including the one adopted for this study, the numerical calculations of the gamma dose rates are three times lower than those from the most commonly-cited reference values for spent LWR fuel used by domestic and international agencies. With only a few publically available experimental data points worldwide for this type of benchmark study, the gamma flux measurements from the Spent Fuel Non Destructive Assay (NDA) measurement campaign can be used to help validate the physics approach used in calculating gamma dose rates. Specifically, this data would validate the gamma flux calculation from spent fuel, which is a key component of these multi-physics techniques for dose rate assessments; dose rate measurements were not part of the NDA campaign. These data are primarily the passive gamma measurements of the spent nuclear fuel assemblies from the Swedish Nuclear Fuel and Waste Management Company (SKB).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Comparison of atmospheric radionuclide dispersion models for a risk-informed consequence-driven advanced reactor licensing framework

Current nuclear facility emergency planning zones (EPZs) are based on outdated distance-based criteria, predating comprehensive dose and risk-informed frameworks. Recent advancements in simulation tools have permitted the development of site-specific, dose, and risk-based consequence-driven assessment frameworks. This study investigated the computation of advanced reactor (AR) EPZs using two atmospheric dispersion models: a straight-line Gaussian plume model (GPM) and a semi-Lagrangian Particle in Cell (PIC). Two case studies were conducted: (1) benchmarking the NRC SOARCA study for the Peach Bottom Nuclear Generating Station and (2) analyzing an advanced INL Heat Pipe Design A microreactor's end-of-cycle inventory. The dose criteria for both cases were 10 mSv at mean weather conditions and 50 mSv at 95th percentile weather conditions at 96 h post-release. Results demonstrated that GPM and PIC estimated similar mean peak dose levels for large boiling water reactors in the farfield case, placing EPZ limits beyond current regulations. For ARs with source terms remaining in the nearfield, PIC modeling without specific nearfield considerations could result in excessively high doses and inaccurate EPZ designations. PIC dispersion demonstrated an order of magnitude higher estimate of nearfield inhalation dose contribution when compared to GPM results. Furthermore, both models significantly reduced EPZ sizing within the nearfield. Thus, reductions in the AR source term may eliminate the need for a separate EPZ.

63 RADIATION, THERMAL, AND OTHER ENVIRON. POLLUTAN↗
DOE OSTI2023

Experiment Design and Preparation for a Shielding Benchmark Utilizing Godiva-IV

An experiment is currently being designed to provide a high-quality shielding benchmark for Criticality Accident Alarm System (CAAS) modeling. Previous benchmarks have suffered from uncertainty primarily due to two factors. The first factor is radiation that is not directly coming from the source, also known as room return. Room return is notoriously difficult to account for in experimental data and to model in benchmarks. The second factor is uncertainty associated with the source term itself. In order to reduce these two sources of uncertainty, this experiment will utilize a room return shield [1] that will reduce the effect of room return and previous work performed to ensure the reproducibility of the source term. The experiment will use the Godiva IV [2] assembly located at the National Criticality Experiments Research Center (NCERC) to provide a neutron source representative of a criticality accident. Previous experiments [3]-[4] have shown that Godiva IV is capable of producing both bursts and steady-state emissions of equal magnitude which will allow for reduced uncertainty in the source term. The room return shield will house and isolate the shielding samples, neutron activation foils, and other detectors. Data collected will provide a comparison of the neutron and gamma dose for shielded and unshielded configurations. This data will be used to validate shielding data used for modeling criticality accident alarm systems. Shielding samples included in this benchmark are C, Fe, NaCl, Pb, high density polyethylene, and SiO 2 . This paper details the design of the room return shield, the experimental plan, challenges, and mitigation strategies.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Ion irradiation testing of 14YWT ODS and HT-9

Ion irradiations were performed to compare a new generation of 14YWT, NFA-2, to the prior generation NFA-1. Irradiations were performed at 450 °C and to a peak dose of 500 dpa. The irradiation condition was chosen as a relevant temperature for scoping for an upcoming JOYO neutron irradiation, and the high dose was chosen to evaluate the exceptional microstructural stability to irradiation induced changes of 14YWT. For comparison, HT-9 was irradiated as the benchmark. Irradiation induced changes to the microstructures were limited for the 14YWT samples and extensive for the HT-9 benchmark.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Extreme γ –Ray Radiation Tolerance of Spectrometer–Grade CsPbBr 3 Perovskite Detectors

The perovskite compound CsPbBr 3 has recently been discovered as a promising room-temperature semiconductor radiation detector, offering an inexpensive and easy-to-manufacture alternative to the current benchmark material Cd 1–x Zn x Te (CZT). The performance of CsPbBr 3 sensors is evaluated under harsh conditions, such as high radiation doses often found in industrial settings and extreme radiation in space. Results show minimal degradation in detector performance after exposure to 1 Mrad of Co-60 gamma radiation, with no significant change to energy resolution or hole mobility and lifetime. Additionally, many of the devices are still functional after being exposed to a 10 Mrad dose over 3 days, and those that do not survive can still be refabricated into working detectors. Notably, these results suggest that the failure mode in these devices is likely related to the interface between the electrode and material and their reaction, or the electrode itself and not the material itself. Overall, the study suggests that CsPbBr 3 has high potential as a reliable and efficient radiation detector in various applications, including those involving extreme fluxes and energies of gamma-ray radiation.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Development of Shielding Benchmarks Using the Godiva IV Assembly

The Nuclear Criticality Safety Program (NCSP) is developing a shielding benchmark using the Godiva IV assembly as a source. Its present status is reviewed herein. Even eight decades into the nuclear era, substantial work remains to develop a database of shielding benchmarks to support future nuclear development. A nuclear simulation is only as good as its supporting data, inputs, and validation basis. Uncertainty in these areas is addressed using conservatism, which adds margin and, occasionally, cost. In many shielding situations, high accuracy is not necessary because additional material is not particularly expensive. After all, 1–2 cm of lead often reduces the gamma ray dose substantially. However, in certain areas, conservatism can add unnecessary cost. These areas include mobile shielding applications such as casks, ships, microreactors, and spacecraft, where weight and, thus, margin is expensive. Although these characteristics are side benefits for NCSP shielding benchmark development, the main driver is enabling more reliable placement of criticality accident alarm systems (CAASs) in nuclear material facilities, such as those dedicated to the production of advanced reactor fuels. CAAS placement relies on more than accurate data and code validation. It also relies on sufficiently accurate materials specifications, geometry specifications, and a well-defined, alarm-producing baseline accident. All these things require tacit knowledge and understanding of the problem being evaluated. Benchmarks can help ensure this understanding.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Comparison of Tritium Dose Calculations from MACCS, UFOTRI, and ETMOD

Tritium exhibits unique environmental behavior because of its potential interactions with water and organic substances. Modeling the environmental consequences of tritium releases can be relatively complex and thus an evaluation of MACCS is needed to understand what updates, if any, are needed in MACCS to account for the behavior of tritium. We examine documented tritium releases and previous benchmarking assessments to perform a model intercomparison between MACCS and state-of-practice tritium-specific codes UFOTRI and ETMOD to quantify the difference between MACCS and state of practice models for assessing tritium consequences. Additionally, information to assist an analyst in judging whether a postulated tritium release is likely to lead to significant doses is provided.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2020

Evaluating 225 Ac and 177 Lu Radioimmunoconjugates against Antibody–Drug Conjugates for Small-Cell Lung Cancer

Interest in the use of 225 Ac for targeted alpha therapies has increased dramatically over the past few years, resulting in a multitude of new isotope production and translational research efforts. However, 225 Ac radioimmunoconjugate (RIC) research is still in its infancy, with most prior experience in hematologic malignancies and only one reported preclinical solid tumor study using 225 Ac RICs. In an effort to compare 225 Ac RICs to other current antibody conjugates, a variety of RICs are tested against intractable small-cell lung cancer (SCLC). Here we directly compare, in vitro and in vivo , two promising candidates of each α or β - category, 225 Ac and 177 Lu, versus pyrrolobenzodiazepine (PBD) nonradioactive benchmarks. The monoclonal antibody constructs are targeted to either delta like 3 protein (DLL3), a recently discovered SCLC target, or CD46 as a positive control. An immunocompromised maximum tolerated dose assay is performed on NOD SCID mice, along with tumor efficacy proof-of-concept studies in vivo . We overview the conjugation techniques required to create serum-stable RICs and characterize and compare in vitro cell killing with RICs conjugated to nonspecific antibodies (huIgG1) with either native or site-specific thiol loci against tumor antigen DLL3-expressing and nonexpressing cell lines. Using patient-derived xenografts of SCLC onto NOD SCID mice, solid tumor growth was controlled throughout 3 weeks before growth appeared, in comparison to PBD conjugate controls. NOD SCID mice showed lengthened survival using 225 Ac compared to 177 Lu RICs, and PBD dimers showed full tumor suppression with nine out of ten mice. The exploration of RICs on a variety of antibody-antigen systems is necessary to direct efforts in cancer research toward promising candidates. However, the anti-DLL3-RIC system with 225 Ac and 177 Lu appears to be not as effective as the anti-DLL3-PBD counterpart in SCLC therapy with matched antibodies and portrays the challenges in both SCLC therapy as well as the specialized utility of RICs in cancer treatment.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2021

Verification and Validation Tests of Gamma Library of MC2-3 for Coupled Neutron and Gamma Heating Calculation

For the accurate assessment of the heat generation rate in fast reactors, the gamma library of MC2 -3 and the MC2 -3 + GAMSOR procedure employing the coupled neutron and gamma heating calculation has been thoroughly verified against Monte Carlo results and validated using the ZPPR-15D gamma dose measurement data. NJOY outputs are post-processed in a consistent way with the NJOY procedure to avoid any missing data or double counting of data. Prompt heating for 379 out of 391 isotopes in the gamma library was verified against MCNP6.2 to within 1% relative error in total heating for most isotopes. For both a simple one-dimensional slab problem representing a sodium cooled fast reactor and the Experimental Breeder Reactor II (EBR-II) benchmark problem, the root-mean-square values of assembly power error were less than 0.5% for fuel assemblies, ~1% in blankets and ~1 to ~3% in reflectors compared to MCNP6.2 results. The most plausible cause for the 3% error in a reflector assembly is believed to be the error in the multigroup neutron cross sections for the reflector assembly. For validation, gamma doses measured with thermoluminiscent dosimeters (TLDs) in the ZPPR-15D experiment were calculated using GAMSOR. Due to the uncertainty in the TLD measurement with regards to the energy deposition of photons and neutrons, the validation data leads to a 12.7% uncertainty on the experimental measurement. With this uncertainty bound, the calculated doses all fell within one standard deviation of the measured value. Combined with the accurate calculation of reaction rate distributions and neutron spectrum measurements, these results indicate good agreement for the neutron and gamma heating calculations that were performed.

coupled neutron and gamma heating↗
DOE OSTI2020

Sulfur Pellets Responses to a Bare and Steel Reflected Pulse of the Oak Ridge National Laboratory Health Physics Research Reactor

The experiments analyzed in this report were conducted at the Health Physics Research Reactor (HPRR), also known as the $\textit{Fast Burst Reactor}$. The reactor was designed and built at Oak Ridge National Laboratory (ORNL) in 1961. The HPRR was an unmoderated, unshielded fast reactor that used highly enriched uranium and molybdenum alloy as fuel. The reactor was initially sent to the Nevada Test Site in 1962, where it was used to evaluate radiation doses received as a result of the Hiroshima and Nagasaki bombings during World War II. A few years later, the reactor was sent back to ORNL to be part of the Dosimetry Application Research (DOSAR) facility shown in Figure 1, which included a reactor building shown on the left (west) of the picture and a control and laboratory building in the upper right corner (northeast). The critical assembly was used for numerous technical studies, including systems calibration, dosimetry, radiobiology of plants and animals, testing of radiation alarms, as well as teaching and training in radiation dosimetry and nuclear engineering. Between 1963 and 1987, the HPRR was operated for thousands of hours, achieving criticality close to 10,000 times and motivating many publications. The HPRR was decommissioned in 1987. The goal of this effort was to use historical data from operation of the HPRR to create a criticality accident alarm system (CAAS) benchmark to be included in the $\textit{International Handbook of Evaluated Criticality Safety Benchmark Experiments}$ (ICSBEP Handbook). A thorough inspection was performed of all available documentation and information available. The most promising experiments that were selected for evaluation were those described in the 1987 ORNL report entitled $\textit{Health Physics Research Reactor Reference Dosimetry}$, ORNL-6240. The report includes reference dosimetry results of the shielded and unshielded configurations of the HPRR after burst operations. Because of changes to the reactor positioning and storage systems that were made in 1985, the previous dosimetry reports became obsolete, and the newly designed experiments were needed to create the HPRR’s adjusted dosimetry data. The various results reported in ORNL-6240 include reference doses and dose equivalents from different conventions at different distances and elevations as determined using the detected neutron fluence and conversion factors. The HPRR neutron fluence was obtained through different methods, including sulfur pellet analysis and threshold detector unit data. Information about the HPRR spectrum was also obtained through Bonner sphere measurements. This benchmark is focused on a part of the measured sulfur fluences reported in Appendix H of ORNL-6240. Standard commercial sulfur pellets were placed at different distances from the HPRR centerline during burst operation and were activated due to the 32 S(n,p) 32 P reaction. The resulting 32 P activity was then measured and the information about the corresponding sulfur fluence and/or neutron dose could be extracted. Many of those measurements have 7 been performed with the HPRR in its bare configuration or with different shields (combinations of Lucite, concrete, steel). All the necessary, precise information about material and/or dimensions of the different shields was not found, so it was decided to focus only on the unshielded and steel-shielded configurations to minimize the benchmark uncertainty. A total of 31 cases (24 unshielded and 7 shielded cases at different positions) of sulfur fluence were selected before evaluation to develop the benchmark.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Advancing the Frontiers of Deep Learning for Low-Dose 3D Cone-Beam CT Reconstruction

X-ray computed tomography (CT) is an important noninvasive medical imaging modality for studying the structural details of internal organs. Image reconstruction in CT is an inverse problem of recovering an object's internal structure from the absorption profile of X-ray beams (sinogram) measured using a detector. The classical variational approach for CT reconstruction minimizes an energy functional using an appropriate iterative algorithm. Motivated by the success of deep learning (DL), researchers have begun to leverage training data and enhanced computing capabilities in recent years to produce high-fidelity reconstructed images. Nonetheless, much of the academic research in DL algorithms for CT has focused primarily on the two-dimensional setting (with simplified forward operators and noise model) for proofs-of-concept, and a comprehensive benchmarking of various classical and data-driven CT reconstruction approaches has not beenundertaken. The key objective of our CT reconstruction grand challenge was to promote methodological advancements for both classical and DL-based approaches for clinical CT with a reasonably accurately simulated 3D CT forward operator and noise model. We have utilized the publicly available LIDC-IDRI dataset and simulated sinograms and FDK images corresponding to two dose levels (clinical- and low-dose, constituting two tracks of the challenge) starting from the normal-dose images as the ground truth. In this paper, we summarize the motivation, context, and results of our challenge, and highlight the future research directions in DL for clinical CT.

X-ray tomography↗
DOE OSTI2023

A fast Monte Carlo cell-by-cell simulation for radiobiological effects in targeted radionuclide therapy using pre-calculated single-particle track standard DNA damage data

Introduction: We developed a new method that drastically speeds up radiobiological Monte Carlo radiation-track-structure (MC-RTS) calculations on a cell-by-cell basis. Methods: The technique is based on random sampling and superposition of single-particle track (SPT) standard DNA damage (SDD) files from a “pre-calculated” data library, constructed using the RTS code TOPAS-nBio, with “time stamps” manually added to incorporate dose-rate effects. This time-stamped SDD file can then be input into MEDRAS, a mechanistic kinetic model that calculates various radiation-induced biological endpoints, such as DNA double-strand breaks (DSBs), misrepairs and chromosomal aberrations, and cell death. As a benchmark validation of the approach, we calculated the predicted energy-dependent DSB yield and the ratio of direct-to-total DNA damage, both of which agreed with published in vitro experimental data. We subsequently applied the method to perform a superfast cell-by-cell simulation of an experimental in vitro system consisting of neuroendocrine tumor cells uniformly incubated with 177 Lu. Results and discussion: The results for residual DSBs, both at 24 and 48 h post-irradiation, are in line with the published literature values. Our work serves as a proof-of-concept demonstration of the feasibility of a cost-effective “in silico clonogenic cell survival assay” for the computational design and development of radiopharmaceuticals and novel radiotherapy treatments more generally.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2024

Design of a high-temperature cell for cobalt-60 irradiations of aqueous solutions with in situ UV–visible spectroscopy

To understand the speciation of solutes in aqueous solutions in high temperature radiation environments, we report the design and fabrication of a custom-built, high temperature (≤300 °C) titanium irradiation cell with in situ optical spectroscopy capabilities, as afforded by coupled fiber optic cables. The wetted surfaces of the 8-inch tall cylindrical cell with 3.5 in. diameter are entirely made of titanium, sapphire, and gold, which are chemically and radiolytically inert. The initial benchmarking results are reported, including the baseline spectrum of deionized water as a function of temperature, the stability of a spectrum over 4 h at 100 °C, and an irradiated Fricke dosimetry solution under ambient irradiator temperature conditions (27.0 ± 0.5 °C). Further, the average gamma radiation dose rate in the cell in its current configuration is 26.1 ± 1.3 Gy min -1 . This cell has application in studying several processes throughout the nuclear fuel cycle, including the reactor coolant behavior.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2023

Pacific Northwest National Laboratory Annual Site Environmental Report for Calendar Year 2022 (Final Report)

Pacific Northwest National Laboratory (PNNL), one of the U.S. Department of Energy (DOE) Office of Science’s 10 national laboratories, provides innovative science and technology development in the areas of energy and the environment, fundamental and computational science, and national security. There are three DOE offices within the Richland area. Two are responsible for the Hanford Site, whereas the Pacific Northwest Site Office (PNSO) oversees PNNL. PNNL prepares this Annual Site Environmental Report to meet the requirements of DOE Order 231.1B, Environmental, Safety and Health Reporting, and DOE Order 458.1, Radiation Protection of the Public and the Environment, assuring that the public is informed of any PNNL-Richland Campus or PNNL-Sequim Campus event that could adversely affect the health and safety of the public, site staff, or the environment. The report provides a synopsis of ongoing environmental management performance and compliance activities for operations that occur on the PNNL-Richland Campus in Richland, Washington, and at the PNNL-Sequim Campus near Sequim, Washington. It describes the location of and background for each facility; addresses compliance with applicable DOE, federal, state, and local regulations, and site-specific permits; documents environmental monitoring efforts and their status; presents potential radiation doses to staff and the public in the surrounding areas; and describes DOE-required data quality assurance methods used for data verification. The ASER summarizes site compliance with federal, state, and local environmental laws, regulations, policies, directives, permits, and Orders, and provides environmental management performance benchmarks and their status to the public, regulatory agencies, community officials, Native American tribes, and public interest groups.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2024

Optimize Electron Beam Energy toward In Situ Imaging of Thick Frozen Bio-Samples with Nanometer Resolution Using MeV-STEM

To optimize electron energy for in situ imaging of large biological samples up to 10 μm in thickness with nanoscale resolutions, we implemented an analytical model based on elastic and inelastic characteristic angles. This model has been benchmarked by Monte Carlo simulations and can be used to predict the transverse beam size broadening as a function of electron energy while the probe beam traverses through the sample. As a result, the optimal choice of the electron beam energy can be realized. In addition, the impact of the dose-limited resolution was analysed. While the sample thickness is less than 10 μm, there exists an optimal electron beam energy below 10 MeV regarding a specific sample thickness. However, for samples thicker than 10 μm, the optimal beam energy is 10 MeV or higher depending on the sample thickness, and the ultimate resolution could become worse with the increase in the sample thickness. Moreover, a MeV-STEM column based on a two-stage lens system can be applied to reduce the beam size from one micron at aperture to one nanometre at the sample with the energy tuning range from 3 to 10 MeV. In conjunction with the state-of-the-art ultralow emittance electron source that we recently implemented, the maximum size of an electron beam when it traverses through an up to 10 μm thick bio-sample can be kept less than 10 nm . This is a critical step toward the in situ imaging of large, thick biological samples with nanometer resolution.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Influence of fluorescent dopants on the vat photopolymerization of acrylate-based plastic scintillators for application in neutron/gamma pulse shape discrimination

Plastic scintillators, a class of solid-state materials used for radiation detection, were additively manufactured with vat photopolymerization. The photopolymer resins consisted of a primary dopant and a secondary dopant dissolved in a bisphenol A ethoxylate diacrylate-based matrix. The absorptive dopants significantly influence important print parameters, for example, secondary dopants decrease the light penetration depth by a factor > 12×. The primary dopant 2,5-diphenyloxazole had minimal impact on the printing process even when loaded at 25 % by mass of the resin. Working curve measurements, which relate energy dose to cure depth, were performed as a function of feature size to further assess the influence of dopants. Photopatterns smaller than 150 µm width had apparent increases in critical energy dose compared to larger photopatterns, while all resins maintained printed features in line gratings with 50 µm of separation. Printed scintillator monoliths were compared to scintillators cast by traditional molding, demonstrating that the layer-by-layer printing process does not decrease scintillation response. In conclusion, a maximum light output of 31 % of a benchmark plastic scintillator (EJ-200) and successful pulse shape discrimination were achieved with 20 % by mass 2,5-diphenyloxazole as the primary dopant and 0.1 % by mass 9,9-dimethyl-2,7-distyrylfluorene as the secondary dopant in printed scintillator samples.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗