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DOE OSTI2023

Development of Spectroscopic Tools for Nuclear Forensics Signatures Development

Diversion of nuclear materialsfrom legitimate commercial processes for use by nefarious actors is a frightening and surreptitious problem. Smuggled nuclear material can be used to manufacture a simple radiological dispersion device (RDD) and potentially incorporated into the development of a nuclear weapon. With some frequency, smuggled nuclear material is interdicted in foreign countries as actors cross border checkpoints. Nuclear material interdicted in this way can be analyzed using a variety of tools. Plutonium-239 is commonly used as the fissile material in nuclear weapon production as an alternative to uranium-235. Plutonium in a production, refining, or finishing facility will exist in many forms including oxide precursors (PuF4, PuF3, Pu oxalate, etc.), oxide, and metal. The ability to identify plutonium in each of these forms and determine properties, characteristics, or history of these materials is crucial for determining if the processes occurring in a nuclear facility are legitimate or not. This presentation will focus on the development and application ofspectroscopic tools developed at SRNL using doubled-walled cells (DWC) to characterize Pu-bearing compounds. Advances in the use of vibrational spectroscopy techniques (Raman and IR), UV-VIS-SWIR diffuse reflectance spectroscopy, gamma emission spectroscopy, luminescence spectroscopy, laser-induced breakdown spectroscopy in conjunction with XRD and SEM in the characterization of Pu-bearing compounds will be presented with emphasis in material processing history.

Villa-Aleman, Eliel↗

DOE OSTI2022

Gamma-ray spectroscopy using angular distribution of Compton scattering

We present a novel method to perform gamma-ray spectroscopy suitable for extreme pileup situations, wherein traditional pulse-height spectroscopy is infeasible or impractical. One example of such application is spectral characterization of intense laser-based radiation sources that employ laser wakefield acceleration, in which a large number of photons can be produced in a well-collimated beam on a picosecond or shorter timescale. The method relies on the angularly-resolved calorimetric measurement of Compton scattering. The probability for a photon to undergo Compton scattering into a given solid angle is a function of its energy and is described by the Klein–Nishina formula. By first Compton scattering the incident beam and then measuring the total energies deposited in multiple detectors placed at various scattering angles, the incident beam energy spectrum can be reconstructed without resolving the individual events in detectors. Here, we compare the performance of an maximum likelihood expectation maximization algorithm with that of an artificial neural network to measure the energy spectrum of a beam of monoenergetic gamma rays from 137Cs decay. We demonstrate that the method can be used to reconstruct the characteristic inverse Compton scattering spectra.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2023

Cumulative fission yields of short-lived fission products from 235 U and 239 Pu measured by HPGe gamma-ray spectroscopy

In this study, we present a preliminary investigation focused on determining cumulative fission yields for short-lived fission products. Our analysis involves examining gamma spectra from the irradiated samples of 235 U and 239 Pu using the High Flux Isotope Reactor. The motivation stems from the observed discrepancy in the antineutrino energy spectrum within the range of 5 to 7 MeV. While several hypotheses have been proposed, a thorough analysis of fission yields provides an additional way of gaining insight into this unexplained phenomenon. Our study suggests that the measured gamma rays from 100 Nb, 140 Cs and 95 Sr are consistent with the expected values. However, 93 Rb, 96 Y, 97 Y and 142 Cs cannot be quantified due to insufficient statistics, interference from other gamma rays and the Compton scattering background. Additionally, the calculated cumulative fission yields based on the measured 140 Cs and 95 Sr are found to be consistent with the JEFF3.3 fission yield library. In conclusion, the present work shows that the potential of improving gamma-ray spectroscopy in the fission yields as a means to improve our understanding of the antineutrino spectrum.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2024

Spectroscopic analysis of Pu-bearing compounds in double-walled cells

Spectroscopic analysis of radiological materials has been historically limited to radiological labs with older or less advanced scientific instrumentation. The development of double-walled cells (DWCs) at the Savannah River National Laboratory (SRNL) has enabled Pu-bearing compounds to be removed from radiological laboratories and studied in our radiologically clean spectroscopy laboratory with state-of-the-art instrumentation. In this manuscript, we discuss the contributions of DWCs that have allowed the application of Raman spectroscopy, diffuse reflectance infrared Fourier transform spectroscopy (DRIFTS), diffuse reflectance spectroscopy (DRS) in the shortwave infrared, and gamma spectroscopy at SRNL. Significant advances have been made in the understanding of the thermal decomposition of Pu(III) and Pu(IV) oxalates, alpha-induced damage to the PuO 2 crystal lattice, and the effect of calcination temperatures on the quality of PuO 2 . These techniques have enabled methods to conduct PuO 2 age dating since last calcination and estimate the calcination temperature with Raman spectroscopy and DRS. Additional spectroscopic information measured with DRIFTS has enabled the observation of the evolution of carbon species with calcination temperature, while gamma spectroscopy provides information on age dating since last purification.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2026

Salt mass determination by 22 Na-based radioactive tracer dilution of a highly radioactive molten salt system for pyroprocessing spent oxide nuclear fuels

To demonstrate the feasibility of the 22 Na-based radioactive tracer dilution (RTD) for determining the total salt mass in a 60-kg oxide reduction (OR) system at Idaho National Laboratory (INL) for pyroprocessing spent oxide nuclear fuels, a LiCl- 22 NaCl tracer salt was prepared and spiked to the OR vessel, and salt samples were analyzed by gamma spectroscopy. Due to the high radioactivity (around 10 mCi/g) of the OR salt, mainly from 137 Cs, traditional gamma spectroscopy methods, which typically involve diluted salt samples, was not effective for analyzing the RTD samples. Since the 22 Na tracer radioactivity was very low, undiluted samples were necessary to detect the small amount of 22 Na tracer. Here, despite the high gamma radioactivity of the OR salt, increasing the detector-sample distance from 20 cm to 40 cm during gamma spectroscopy allowed the detection of radioactivity as low as 0.12 µCi/g 22 Na. The preparation of the 22 NaCl-LiCl tracer salt suitable for spiking into the OR salt was successful, and the mass determined by RTD was consistent with that estimated by salt level measurement. Therefore, the 22 Na-based RTD technique for total mass determination of the 60-kg molten salt system for OR is feasible, though more development work, particularly in the improving the uncertainty, is needed .

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Thallium-based scintillators for high-resolution gamma-ray spectroscopy: Ce 3+ - doped Tl 2 LaCl 5 and Tl 2 LaBr 5

In this paper we report on the crystallographic and scintillation properties of Tl 2 LaCl 5 :Ce 3 + and Tl 2 LaBr 5 :Ce 3 + , two novel thallium-containing high-resolution scintillators for gamma-ray spectroscopy. Crystals of Tl 2 LaCl 5 :Ce 3 + and Tl 2 LaBr 5 :Ce 3 + were grown by the Vertical Bridgman method up to 1-inch diameter and 1-inch long. Single crystals of Tl 2 LaCl 5 :Ce 3 + and Tl 2 LaBr 5 :Ce 3 + belong to the orthorhombic system with space group 62 and have a density of 5.16 and 5.98 g/cm 3 , respectively. The scintillators show high light yields of up to 68,000 photons/MeV, excellent gamma-ray energy resolution of ≤ 3% at 662 keV, a fast scintillation decay, and a proportional response over a wide range of energies from 32 keV up to 1275 keV. Density Functional Theory calculations show that the Ce 3 + energy levels are inside the bandgap despite the smaller bandgap of Tl 2 LaCl 5 and Tl 2 LaBr 5 compared to K 2 LaCl 5 and K 2 LaBr 5 . Finally, a systematic Ce 3 + concentration study was performed for Tl 2 LaCl 5 :Ce 3 + and trends observed.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2020

Results for Gamma-Ray Spectroscopy Characterization of Residual Radioactivity on Personal Protective Clothing

In early 2019, Health Physics Services (HPS) personnel put together a test plan to use the shielded gamma-ray spectroscopy instruments in the Health Physics Services Technical Laboratory (HPSTL) to perform a series of measurements on laundered personal protective clothing used in radiological areas at the Savannah River Site (SRS). That test plan describes the methodology used to determine the quantities of laundry to be obtained from operational facilities, the methods for quality control checking the instruments throughout the measurement period, and the physical arrangement of samples in the counting shields. The test plan is documented in SRNS-J6700-2019-00037 (Reference 1).

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2021

Holistic digital gamma-ray spectroscopy methods and instrumentation for high-throughput high-resolution applications

Method of real-time adaptive digital pulse signal processing for high count rate gamma-ray spectroscopy applications includes receiving a preamplifier signal at a pulse deconvolver, the preamplifier signal including resolution deterioration resulting from pulse pile-up. The method further includes generating a deconvoluted signal, by the pulse deconvolver, from the preamplifier signal, the deconvoluted signal having less resolution deterioration as compared to the received preamplifier signal. The method furthermore includes shaping of the deconvoluted signal by a trapezoid filter, the shaping comprising adjusting a shaping parameter of the trapezoid filter for an incoming signal based on a time separation from a subsequent incoming signal.

Hawari, Ayman I.↗

DOE OSTI2020

Replacing photomultiplier tubes with silicon photomultipliers for nuclear safeguards applications

Photomultiplier tubes (PMTs) have been used for decades as the default light detection technology for scintillator-based radiation monitors. PMTs come with a handful of disadvantages, however, including large volume usage, fragility, high-voltage requirements, and susceptibility to magnetic fields. Arrays of silicon photomultipliers (SiPMs) are a possible alternative to PMTs providing similar performance while offering improvements in the areas listed. We focus on straightforward “drop-in replacement” evaluations by imposing a single channel output of the SiPM signal and a backend electronics data acquisition and analysis routine identical to that used for PMTs. In the realm of nuclear safeguards, the primary performance metric for gamma spectroscopy is detector resolution, and for neutron detection it is pulse shape discrimination (PSD) to separate neutron and gamma signals. In this work we present the results of replacing a PMT with a 2” x 2” SiPM array mounted, in turn, to the same 2” x 2” cylindrical sodium iodide (NaI) crystal. The ultimate comparison for this gamma spectrometry is confirmation of uranium enrichment standards, ranging from depleted uranium to 93% highly enriched uranium (HEU), by analyzing the resulting spectra with both a simple integral scaling in key energy regions of the spectrum, as well as with the NaIGEM software package. In addition to the gamma spectroscopy work, we will present the latest results on comparing the PSD capabilities of a 4” x 4” SiPM array to a 5” PMT. As with the gamma spectroscopy evaluations, the light detectors are mounted in turn to the same liter-scale organic scintillators. This neutron-focused work includes evaluation of prototype SiPM readout boards with a single output signal summed over 256 individual SiPM pixels, and evident tradeoffs between PSD capability and fast response.

42 ENGINEERING↗

DOE OSTI2020

Replacing photomultiplier tubes with silicon photomultipliers for nuclear safeguards applications

Engineering - Instrumentation related to nuclear s↗

DOE OSTI2022

An approach towards plastic scintillators from thermally activated delayed fluorescent dyes and cross-linkable bismuth compounds

Cross-linked polyvinyltoluene (PVT) based plastic scintillators loaded with dyes exhibiting thermally activated delayed fluorescence (TADF) and various amounts of (caproate)di-(methacrylate) bismuth (CMB) were developed. The co-polymerization of vinyl toluene monomer loaded with different weight ratios of CMB and 1 wt% of TADF dye yielded highly transparent cross-linked plastic scintillators with intense green photoluminescence. The thermal and mechanical properties of the scintillators were investigated using thermogravimetric analysis (TGA), differential scanning calorimetry (DSC) and micro indentation techniques. Here, their performance for radiation detection was assessed by performing gamma spectroscopy using a Cs-137 source and benchmarked to commercial plastic scintillators. The loading of CMB was guided by Monte Carlo N-Particle (MCNP) simulations. The results of this study show that the cross-linking approach and the use of TADF dyes yield scintillators with good mechanical properties with a uniaxial yield strength up to ~66 MPa for compositions loaded with 40 wt% of CMB and high optical quality. However, the light yield in gamma spectroscopy experiments is found to be reduced with the addition of CMB, suggesting a trade-off between mechanical yield strength and light yield in gamma spectroscopy.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2021

Modeling 233 Pa Generation in Thorium-fueled Reactors for Safeguards

Thorium has been considered as a possible alternative to uranium for nuclear fuel for many decades. It is three to four times more abundant in the earth than uranium and produces significantly less long-lived transuranic nuclear waste. Some claim thorium poses fewer proliferation concerns than other fuel types largely due to 232 U buildup (and associated high energy gamma-emitting decay products) in the irradiated thorium fuel. However, to fully explore potential proliferation concerns, generation and subsequent decay of 233 Pa produced in the reactor core still must be studied. With its half-life of 27 days, 233 Pa decays to 233 U, which is an International Atomic Energy Agency (IAEA) defined special fissionable material that can be used for nuclear weapons production. With more research being dedicated to thorium-fueled reactors, and several of these reactor designs possessing online fuel processing (allowing for on-site protactinium separation), it is important to understand this potential proliferation pathway. In particular, it is theoretically possible to extract protactinium from the irradiated fuel salt before it decays into 233 U. This hypothetical potential diversion can become an even greater proliferation concern if the extracted protactinium is purified through a second separation of protactinium approximately ten days later to remove the short half-life decay products of 232 Pa and 234 Pa, thus resulting in a higher concentration of the 233 Pa isotope, which decays into weapons usable 233 U with hardly any 232 U or 234 U in it. To estimate the concern of this potential proliferation challenge of thorium, different nuclear material accountancy techniques were reviewed for their viability to quantify 233 Pa if extracted from used thorium fuel. Characteristics of interest included technology maturity, cost, precision, and time taken to acquire results. Some technologies, like hybrid K-edge densitometry and passive gamma spectroscopy, appear to be viable techniques based on current literature. Due to the limited scope of this project, only passive gamma spectroscopy was further investigated. Three different reactor types (PWR, CANDU, MSR) were modeled with mixed thorium-uranium oxide fuels that were burned until the fuel was spent. The protactinium in the used fuel was extracted at the time of shutdown and the change in isotopic content of the protactinium quantified. Gamma spectroscopy simulations were performed for the protactinium isotopes and their decay products at various decay times. Given the simplicity of the models and large assumptions made (e.g. no background, no shielding, no self-attenuation), the initial results indicate that though 233 Pa is detectible for all the reactor types modeled at all decay times (0 to 300 days), more work should be done with higher fidelity models.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2023

Triple-mode detector for fast neutrons, slow neutrons, and gamma ray spectroscopy

This disclosure provides systems, methods, and apparatus related to neutron detection and gamma ray detection. In one aspect, a detector comprises a scintillator structure that comprises an organic scintillator and an inorganic scintillator. The organic scintillator is in the form of one or more elements of a specified length. The inorganic scintillator is in the form of one or more elements of the specified length. First ends of the one or more organic scintillator elements and first ends of the one or more inorganic scintillator elements define a first surface. Second ends of the one or more organic scintillator elements and second ends of the one or more inorganic scintillator elements define a second surface.

Derenzo, Stephen↗

DOE OSTI2021

Multi-pulse time resolved gamma ray spectroscopy of the advanced radiographic capability using gas Cherenkov diagnostics

The advanced radiographic capability located at the National Ignition Facility (NIF) uses high intensity, short pulse lasers to create bright photon sources for diagnosing high energy density experiments. There are radiographic needs for a multi-frame time-resolved MeV gamma diagnostic for experiments on the NIF with sub-nanosecond resolution. A series of experiments demonstrated measurements of MeV x-ray spectra resolved with a time separation of a few nanoseconds through the use of gas Cherenkov detectors. In this work, a two-pulse radiographic experiment found a 30% reduction in > 2.8 MeV photon flux compared to the first frame exposure.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2024

Radiation Monitoring for Volatilized Zinc Contamination Using Gamma-Ray Imaging and Spectroscopy

Gamma-ray imaging is a tool that has grown in importance in the applications of non-destructive assay (NDA) for radioactive survey and analysis of nuclear facilities. Imaging techniques have shown great promise in providing valuable information involving radioactive waste management and contamination prevention. For the application studied in this work, 65 Zn has been identified as a radioactive contaminant during tritium extraction. Due to the volatile nature of 65 Zn under the pressure and temperature changes during extraction operations, 65 Zn can easily travel through components of the extraction system as vapor, making it difficult to trap. Previous research involving the development of a filtration system showed that the 65 Zn can be trapped, mitigating product contamination. However, during the extraction process, direct analysis of the equipment to confirm that zinc contamination is trapped in the filter and has not spread to other components is impractical. Here, in this situation, the need to assay the location of the contamination with little-to-no interference with operations is vital. In this work, we demonstrate the use of a commercialized 3D position-sensitive CdZnTe (CZT) gamma-ray imaging spectrometer to provide analysis of the 65 Zn contamination. Onsite measurements during an extraction process are studied to assess the location and migration of the 65 Zn. The results obtained from real-time glovebox monitoring demonstrate the feasibility of gamma-ray imaging for localizing the contamination and providing a preliminary qualitative assessment that is intended to be used in future work quantifying the contamination build-up and activity over time.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2022

Discrimination of phosgene from cyanogen chloride in recovered chemical munitions using tagged neutron interrogation with high-resolution gamma-ray spectroscopy

The Associated Particle technique has recently been utilized with a germanium-based high-resolution $\gamma$-ray spectrometer and assessed for its capability to improve field identification of recovered chemical warfare (CW) materiel through prompt gamma-ray neutron activation analysis (PGNAA) measurements. An accurate chemical identification is necessary for the safe handling, destruction, and disposal of such hazardous items. A particularly challenging pair of CW agents commonly found in recovered munitions are phosgene (CG) and cyanogen chloride (CK), which have two of three elements in common, i.e. chlorine and carbon, but differ in the third being either oxygen or nitrogen. The detection of both latter elements is complicated by high oxygen concentration in the field environment which interferes with the small signal produced from the chemical agents. The matter is further complicated by the precautionary field practice of overpacking recovered munitions with vermiculite in larger steel multiple round containers (MRCs), which places additional oxygen-rich material around the munition while further attenuating an already weak signal emitted from the munition center. Here, this work reports quantitative results from realistic field measurements of CG and CK simulants in mock 4.2-inch mortar rounds overpacked with vermiculite in a large MRC. Results obtained with the AP technique are compared to those obtained with the traditional PGNAA approach for both overpacked- and bare-munition measurements. The AP technique is shown to provide a much more confident discrimination between the two chemicals, particularly for the more challenging field-relevant overpacked measurements, where a significant gain in sensitivity to all the key elements chlorine, carbon, nitrogen and oxygen is achieved.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2024

Evaluation of material accountancy techniques for 233 Pa from thorium nuclear fuels

Thorium is a promising alternative to uranium as nuclear fuel with advantages such as higher abundance, lower production of long-lived transuranic elements, and potentially better proliferation resistance. However, thorium presents a potential pathway for proliferation where produced 233 Pa can be diverted for the clandestine production of safeguarded 233 U. To prevent this, the ability to detect and measure 233 Pa must be assessed. This paper reviews several nuclear material accountancy techniques to determine their suitability for detecting 233 Pa extracted from irradiated thorium fuel. Hybrid K-edge densitometry and passive gamma spectroscopy have been found to be the best options based on technology maturity, cost, accuracy, and acquisition time. Thorium can be used in various reactor designs such as pressurized water reactors (PWRs), Canada deuterium uranium (CANDU) reactors, and molten salt reactors (MSRs). Therefore, thorium-uranium oxide fueling was modeled for three representative reactors (PWR, CANDU, MSR), burning the fuel to 47 GWd/MTHM for PWR, 19 GWd/MTHM for CANDU, and at a steady power of 52.711 MW/MTHM for MSR. Within each model, the protactinium element in the used fuel was extracted and its isotopic content analyzed. Simulated results indicated that 233 Pa can be detected using passive gamma spectroscopy in each fuel type at all decay times (0–300 days) following separation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Testing, Calibration, and UxS Integration of the Kromek GR1 Plus CZT Gamma Spectrometer

Collecting radiation measurements can be a hazardous task, especially in the presence of highly active sources. Various scenarios necessitate source search, classification, and quantification, often with limited or no a priori information. Activities such as disaster mitigation, emergency response, environmental monitoring, and site remediation may involve dangerous radioactive sources. In these situations, there is a pressing need for remote monitoring capabilities that protect human operators from potential harm and enhance adherence to the principle of “As Low as Reasonably Achievable” (ALARA) for radiation doses. The first step toward achieving remote radiation measurement capabilities is the remote operation of a radiation sensor. Once this milestone is reached, the next challenge is to integrate this remote sensing capability into suitable actuation agents, collectively referred to as uncrewed systems (UxS). These systems include familiar platforms such as robotic quadrupeds, aerial multirotor vehicles, and ground vehicles, any of which may be teleoperated, act autonomously, or utilize a combination of both. A critical factor in achieving remote radiation sensing is the availability of data from the appropriate sensor. Many commercially available radiation sensors have closed-source documentation for their communication protocols. Typical end-user products are often self-contained, handheld devices designed for manual measurement scenarios. While there are commercial off-the-shelf (COTS) integrations of radiation sensors with UxS available for purchase, these solutions are typically tailored for specific use cases and may not meet the requirements of different applications. This paper discusses efforts to remotely acquire radiation measurements from a small form-factor CZT gamma spectrometer. Sandia National Laboratories has successfully demonstrated the initial capability to integrate low size, weight, and power (SWaP) gamma spectroscopy into various UxS, alongside co-located GPS data logging and sensor calibration and qualification. With remote gamma spectroscopy achieved, the stage is set for UxS integration of this capability.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗