Search NASA⌕ Search

SEARCH · Search NASA

Results for “Minimum Detectable Activities”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 55 records · Page 3

Characterization of a large-area iQID imager for safeguards applications

Environmental sampling and sample analyses by the International Atomic Energy Agency is a key method for verifying declarations in international safeguards agreements. Quantitative Digital Autoradiography for Environmental Samples is a system developed by Pacific Northwest National Laboratory to examine environmental swipe samples (ESS) prior to subsequent, destructive analysis. This approach is passive and noninvasive for imaging beta and gamma emissions, which transmit through the plastic bags containing the ESS. Alpha particle imaging is also possible with different ESS containers. This technology is based on a portable, real-time, event-counting, high spatial resolution system called iQID (ionizing-radiation Quantum Imaging Detector), which is made from commercial, off-the-shelf components. This includes a scintillator in direct contact with a micro-channel plate image intensifier and a lens for imaging the intensified scintillation events onto a CCD/CMOS camera. The QDARES large-area iQID includes a fiber-optic taper coupled to the image intensifier which increases the imaging area to a ~100 cm square with a 105 mm x 105 mm taper. In this study, we describe new characterization results, including spatial resolution, energy response, minimum detectable activity for several isotopes, detector uniformity, and count-rate capability.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

Monte Carlo investigation of a high-efficiency, two-plane Compton camera for long-range localization of radioactive materials

Compton cameras have become an instrument of interest for long-range localization of radioactive materials in nuclear-nonproliferation applications. In this work, a specialized simulation tool was developed for the optimization of a Compton camera through realistic Monte Carlo simulations. This tool can be used for Compton cameras with different geometries. The MCNPX-PoliMi code was used to simulate Compton scatters and photoelectric absorptions in the camera's detectors. The imaging capability is derived from the physics of Compton scattering and the full-energy information obtained in photoelectric absorptions. The simulated system is a 1 1 m2 Compton camera consisting of two planar arrays of photon detectors. Several scintillators were evaluated: LaBr3, CaF2, and NaI(Tl) (all inorganic scintillators), and C9H10 (plastic organic scintillator). The investigation was carried out with a 137Cs source. A minimum detectable activity (MDA) was defined and used to assess the performance of cameras based on different detectors. The simulation results show that C9H10 is a reliable, low-cost, scatter-plane material, increasing the MDA by ~0.2 mCi when compared to a CaF2 scatter plane. On the other hand, the intrinsic background of LaBr3 undermines its above-average energy resolution as the MDA increased by ~0.5 mCi when compared to a NaI(Tl) absorption plane.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Completion Report for Well ER-EC-11 Corrective Action Units 101 and 102: Central and Western Pahute Mesa

Well ER-EC-11 was drilled for the U.S. Department of Energy, National Nuclear Security Administration Nevada Site Office, in support of the Nevada Environmental Restoration Project at the Nevada National Security Site (formerly Nevada Test Site), Nye County, Nevada. The well was drilled in September and October 2009 as part of the Pahute Mesa Phase II drilling program. A main objective was to investigate radionuclide migration down-gradient from Well Cluster ER-20-5 and Well ER-20-7 and across the northern Timber Mountain moat structural zone into the area referred to as the Bench, between Pahute Mesa and the Timber Mountain caldera complex. A secondary purpose of the well was to provide detailed hydrogeologic information for the shallow- to intermediate-depth Tertiary volcanic section in the Bench area. This well also provided detailed hydrogeologic information in the Tertiary volcanic section to reduce uncertainties within the Pahute Mesa-Oasis Valley hydrostratigraphic framework model (Bechtel Nevada, 2002). The main 52.1-centimeter hole was drilled to a depth of 507.5 meters and then opened to a diameter of 66.0 centimeters. It was cased with 50.8-centimeter casing to 504.9 meters. The hole diameter was then decreased to 47.0 centimeters, and drilling continued to a total depth of 979.3 meters. It was then cased with 34.0-centimeter casing set at 965.5 meters. The hole diameter was then decreased to 31.1 centimeters and the borehole was drilled to a total depth of 1,264.3 meters. The completion casing string, set to the depth of 1,262.5 meters, consists of 19.4-centimeter stainless-steel casing hanging from 19.4-centimeter carbon-steel casing. The stainless-steel casing has two slotted intervals open to the Tiva Canyon and Topopah Spring aquifers. Four piezometer strings were installed in Well ER-EC-11. A string of carbon-steel 6.0-centimeter tubing with one slotted interval was inserted outside the 50.8-centimeter casing, within the 66.0-centimeter borehole for access to the Timber Mountain aquifer, and landed at 475.3 meters. A second string of 6.0-centimeter tubing with one slotted interval was inserted outside the 34.0-centimeter casing, within the 47.0-centimeter borehole for access to the Benham aquifer, and landed at 911.7 meters. A third piezometer string consists of 7.3-centimeter stainless-steel tubing that hangs from 6.0-centimeter carbon-steel tubing via a crossover sub. This string was landed at 1,029.5 meters to monitor the Tiva Canyon aquifer. The deepest string of 7.3-centimeter tubing was landed at 1,247.8 meters to monitor the Topopah Spring aquifer. Data collected during and shortly after hole construction include composite drill cuttings samples collected every 3.0 meters, 67 percussion gun and rotary sidewall core samples, various geophysical logs, fluid samples (for groundwater chemistry analysis and tritium measurements), and water-level measurements. The well penetrated 1,264.3 meters of Tertiary volcanic rock, including three saturated welded-tuff aquifers and one saturated lava-flow aquifer. A water level was measured in the Timber Mountain aquifer at 449.6 meters, during open-hole geophysical logging on September 20, 2009. The fluid level measured after the total depth was reached and the upper aquifer was cased off was 450.0 meters when measured in the open borehole on October 17, 2009. Measurements on samples taken from the undeveloped well indicated that tritium levels averaging approximately 12,430 picocuries per liter (less than Safe Drinking Water Act levels) were encountered within the Benham aquifer. Tritium was below the minimum detectable activity concentration for samples collected from the Tiva Canyon aquifer and the Topopah Spring aquifer.

54 ENVIRONMENTAL SCIENCES↗

The iSERIES™ radon progeny compensation algorithm and its application to air filters

Sandia has developed methods to perform a rapid count using Canberra’s iSERIES™ Alpha/Beta Counting System, the iSOLO™. The purpose of this report is to show that by using the iSERIES™ built-in radon progeny compensation algorithm, the radon progeny-free gross alpha air concentration can be determined for a sample taken using a high volume air sampler or a lapel device within a few hours post-sampling. Also, a special form of the Currie Minimum Detectable Activity estimation was derived for use with the radon progeny stripping algorithm after it was discovered that the expression used by the Canberra software did not take the counts due to radon progeny into consideration as part of the background. This rapid air sample analysis proves to be a vast improvement to the current method that has an average turn-around time between seven and ten days.

Fournier, Sean Donovan↗

Comparison of Results for Quarter 1 Surface Water Split Samples Collected at the Nuclear Fuel Services Site, Erwin, Tennessee

Oak Ridge Associated Universities (ORAU), under the Oak Ridge Institute for Science and Education (ORISE) contract, collected split surface water samples with Nuclear Fuel Services (NFS) representatives on August 22, 2012. Representatives from the U.S. Nuclear Regulatory Commission and Tennessee Department of Environment and Conservation were also in attendance. Samples were collected at four surface water stations, as required in the approved Request for Technical Assistance number 11-018. These stations included Nolichucky River upstream (NRU), Nolichucky River downstream (NRD), Martin Creek upstream (MCU), and Martin Creek downstream (MCD). Both ORAU and NFS performed gross alpha and gross beta analyses. The comparison of results using the duplicate error ratio (DER), also known as the normalized absolute difference. A DER ≤ 3 indicates that, at a 99% confidence interval, split sample results do not differ significantly when compared to their respective one standard deviation (sigma) uncertainty. The NFS split sample report does not specify the confidence level of reported uncertainties. Therefore, standard two sigma reporting is assumed and uncertainty values were divided by 1.96. A comparison of split sample results, using the DER equation, indicates one set with a DER greater than 3. A DER of 3.1 is calculated for gross alpha results from ORAU sample 5198W0003 and NFS sample MCU-310212003. The ORAU result is 0.98 ± 0.30 pCi/L (value ± 2 sigma) compared to the NFS result of -0.08 ± 0.60 pCi/L. Relatively high DER values are not unexpected for low (e.g., background) analyte concentrations analyzed by separate laboratories, as is the case here. It is noted, however, NFS uncertainties are at least twice the ORAU uncertainties, which contributes to the elevated DER value. Differences in ORAU and NFS minimum detectable activities are even more pronounced. comparison of ORAU and NFS split samples produces reasonably consistent results for low (e.g., background) concentrations.

54 ENVIRONMENTAL SCIENCES↗

Quantitative Assessment of Detection Frequency for the INL Ambient Air Monitoring Network

A quantitative assessment of the Idaho National Laboratory (INL) air monitoring network was performed using frequency of detection as the performance metric. The INL air monitoring network consists of 37 low-volume air samplers in 31 different locations. Twenty of the samplers are located on INL (onsite) and 17 are located off INL (offsite). Detection frequencies were calculated using both BEA and ESER laboratory minimum detectable activity (MDA) levels. The CALPUFF Lagrangian puff dispersion model, coupled with 1 year of meteorological data, was used to calculate time-integrated concentrations at sampler locations for a 1-hour release of unit activity (1 Ci) for every hour of the year. The unit-activity time-integrated concentration (TICu) values were calculated at all samplers for releases from eight INL facilities. The TICu values were then scaled and integrated for a given release quantity and release duration. All facilities modeled a ground-level release emanating either from the center of the facility or at a point where significant emissions are possible. In addition to ground-level releases, three existing stacks at the Advanced Test Reactor Complex, Idaho Nuclear Technology and Engineering Center, and Material and Fuels Complex were also modeled. Meteorological data from the 35 stations comprising the INL Mesonet network, data from the Idaho Falls Regional airport, upper air data from the Boise airport, and three-dimensional gridded data from the weather research forecasting model were used for modeling. Three representative radionuclides identified as key radionuclides in INL’s annual National Emission Standards for Hazardous Air Pollutants evaluations were considered for the frequency of detection analysis: Cs-137 (beta-gamma emitter), Pu-239 (alpha emitter), and Sr-90 (beta emitter). Source-specific release quantities were calculated for each radionuclide, such that the maximum inhalation dose at any publicly accessible sampler or the National Emission Standards for Hazardous Air Pollutants maximum exposed individual location (i.e., Frenchman’s Cabin) was no more than 0.1 mrem yr–1 (i.e., 1% of the 10 mrem yr–1 standard). Detection frequencies were calculated separately for the onsite and offsite monitoring network. As expected, detection frequencies were generally less for the offsite sampling network compared to the onsite network. Overall, the monitoring network is very effective at detecting the potential releases of Cs-137 or Sr-90 from all sources/facilities using either the ESER or BEA MDAs. The network was less effective at detecting releases of Pu-239. Maximum detection frequencies for Pu-239 using ESER MDAs ranged from 27.4 to 100% for onsite samplers and 3 to 80% for offsite samplers. Using BEA MDAs, the maximum detection frequencies for Pu-239 ranged from 2.1 to 100% for onsite samplers and 0 to 5.9% for offsite samplers. The only release that was not detected by any of the samplers under any conditions was a release of Pu-239 from the Idaho Nuclear Technology and Engineering Center main stack (CPP-708). The methodology described in this report could be used to improve sampler placement and detection frequency, provided clear performance objectives are defined.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Measurements Conducted on an Unknown Object Labeled Pu-239

Measurements were carried out on 12 November 2013 to determine whether Pu-239 was present on an object discovered in a plastic bag with label “Pu-­239 6 uCi.” Following initial survey measurements to verify that the object was not leaking or contaminated, spectra were collected with a High Purity Germanium (HPGe) detector with object positioned in two different configurations. Analysis of the spectra did not yield any direct evidence of Pu-­239. From the measured spectra, minimum detectable activity (MDA) was determined to be approximately 2 uCi for the gamma-­ray measurements. Although there was no direct evidence of Pu-239, a peak at 60 keV characteristic of Am-­241 decay was observed. Since it is very likely that Am-­241 would be present in aged plutonium samples, this was interpreted as indirect evidence for the presence of plutonium on the object. Analysis of this peak led to an estimated Pu-­239 activity of 0.02–0.04 uCi, or <1x10 -6 grams.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Uncertainty Quantification for Nuclear Safeguards and Non-Destructive Assay - Fact Sheet

UQ is the scientific art of generating confidence statements. Without defensible UQ physical measurements and calculations have no meaning. The UQ team performed a series of case studies with the following results: minimum detectable activity of a Tomographic Gamma-Ray Scanning system was determined using the Currie formalism. Bootstrapping method was used to generate fluorescence yield parameters and uncertainties. For Hybrid K-Edge Densitometry (HKED) measurements, these parameters can be used to predict the concentration of plutonium and quantify uncertainty in HKED models for nuclear safeguards measurements. Detection efficiency was determined for a coincidence counter using covariance data. The use of covariance information drastically reduced the total uncertainty in the average detection efficiency.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

Characterization of a Commercial Silicon Beta Cell

Silicon detectors are of interest for the verification of the Comprehensive Nuclear-Test-Ban Treaty (CTBT) due to their enhanced energy resolution compared to plastic scintillators beta cells. Previous work developing a figure-of-merit (FOM) for comparison of beta cells suggests that the minimum detectable activity (MDA) could be reduced by a factor of two to three with the use of silicon detectors. Silicon beta cells have been developed by CEA (France) and Lares Ltd. (Russia), with the PIPSBox developed by CEA being commercially available from Canberra for approximately $35k, but there is still uncertainty about the reproducibility of the capabilities in the field. PNNL is developing a high-resolution beta-gamma detector system in the shallow underground laboratory, which will utilize and characterize the operation of the PIPSBox detector. Throughout this report, we examine the capabilities of the PIPSBox as developed by CEA. The lessons learned through the testing and use of the PIPSBox will allow PNNL to strategically develop a silicon detector optimized to better suit the communities needs in the future.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

A Figure-of-Merit for Beta Cell Detector Characterization

In order to decrease the minimum detectable activities (MDAs) of beta-gamma radioxenon detectors, it is important to increase the ability to resolve the individual isotopes. One proposed method for doing this is to increase the energy resolution of the beta cell through the use of silicon detectors. While silicon detectors can improve the energy resolution, it is accompanied with a decrease in detection efficiency compared to plastic scintillator beta cells. Due to the uncertainty on the impact of the competing variables, we have developed a figure-of-merit (FOM) capable of determining the impact of detector parameters on the MDAs. By utilizing the FOM to analyze different detectors, we are able to directly compare current and future detectors and estimate their impact on the radioxenon MDAs.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Project 57 Air Monitoring Report: January 1 through December 31, 2017

On April 24, 1957, the Atomic Energy Commission (AEC) (now the Department of Energy [DOE]) conducted the Project 57 safety experiment in western Emigrant Valley northeast of the Nevada National Security Site (NNSS, formerly the Nevada Test Site) on lands withdrawn by the Department of Defense (DOD) for the Nevada Test and Training Range (NTTR). The test was undertaken to develop (1) a means of estimating plutonium distribution resulting from a non-nuclear detonation; (2) biomedical evaluation techniques for use in plutonium-laden environments; (3) methods of surface decontamination; and (4) instruments and field procedures for prompt estimation of alpha contamination (Shreve, 1958). Although the test did not result in the fission of nuclear materials, it did disseminate plutonium across the land surface. Following the experiment, the AEC fenced the contaminated area and returned control of the surrounding land to the DOD. Various radiological surveys were performed in the area and in 2007, the DOE expanded the demarked Contamination Area (CA) by posting signs 200 to 400 ft (60 to 120 m) outside of the original fence. Plutonium in soil attaches preferentially to smaller particles (Tamura, 1985; Friesen, 1992; Murarik et al., 1992; and Misra et al., 1993). Therefore, redistribution of soil particles by wind (dust) and water are the mechanisms most likely to transport plutonium beyond the boundary of the Project 57 CA. Monitoring was implemented in 2011 by Desert Research Institute (DRI) to determine if radionuclide contamination was detectable in samples of airborne dust and to characterize meteorological and environmental parameters that influence dust transport. Initially, two monitoring stations consisting of radiological, meteorological, and dust sampling equipment were installed near the southeast and northeast corners of the CA. In January 2015, the original monitoring stations were dismantled and moved farther to the west along the CA boundary. This move was made to place the monitoring stations downwind of ground zero and the High Contamination Area (HCA) during the dominant northerly and southerly winds. Samples of particles suspended in the air are collected every two weeks and submitted for laboratory assessments of gross alpha and gross beta radioactivity, and for determinations of gamma-emitting radionuclides. In calendar year (CY) 2017, the mean gross alpha concentration at monitoring station P57-4 was approximately 4.25 times the value at monitoring station P57-3. In CY2015 and CY2016, the difference in mean gross alpha concentration at the two stations was approximately 1.5. The mean gross alpha concentrations at P57-3 and P57-4 were approximately 1.9 and 7.9 times, respectively, the mean gross alpha concentrations at the surrounding Community Environmental Monitoring Program (CEMP) stations. The minimum gross alpha concentrations reported for the Project 57 stations are in the range of the minimum values observed at the CEMP stations. However, maximum gross alpha concentrations are greater than the maximum concentrations observed at the surrounding CEMP stations. Gamma spectroscopy analyses resulted in the detection of seven radionuclides. Except for americium-241 (Am-241), all reported radionuclides occur naturally in the environment. The Am-241 detection by gamma spectroscopy was confirmed by alpha spectroscopy. During each quarter of CY2017, four suspended particulate matter samples were selected for alpha spectroscopy analysis. Two were selected each quarter from the regular biweekly samples collected at each of the two monitoring stations. Of the eight samples from P57-3, one was reported to have plutonium-238 (Pu-238) above the minimum detectable activity (MDA) and five were reported to have plutonium 239/240 (Pu-239/240) above the MDA. Of the eight samples from P57-4, three were reported to have Pu-238 above the MDA and six were reported to have Pu-239/240 above the MDA. Soil material is also transported by saltation, which is a wind driven phenomena that bounces sand-sized soil particles across the land surface because they are too heavy to be suspended in the air. Samples of particles transported by saltation were collected downwind and upwind of the CA at both monitoring stations in May 2017 after a 190-day deployment and again in September 2017 after a 114-day deployment. Mass of the saltation samples was determined after oven drying. The September sample had greater mass than the May sample. The same relationship was observed for the suspended particulate matter samples during periods of saltation sample collection. The difference in the collected mass appears to be because of the amount of soil moisture. Higher soil moisture exerts additional isotopic bonds between soil particles and water molecules, and between water molecules that envelop soil particles, making it more difficult for wind to raise the particles as dust. In general, samples collected from saltation traps facing south contained more material than samples collected from traps facing north. This suggests that although saltation material may have been moving back and forth in conjunction with the two dominant wind directions, there was a net trend for saltation material to be transported toward the north. Radiological analyses of the saltation samples show that the smaller particle size faction (<63 μm) generally produced the highest activity levels of Am-241, Pu-238, and Pu-239/240. In general, the isotope activity levels decreased as the particle size increased. These results are consistent with the expectation that the greatest potential for transport of radiological contamination is with the smaller particles. Thermoluminescent dosimeters (TLDs) indicated that the average annual external radioactivity dose at the monitoring stations is slightly higher than the dose determined at the surrounding CEMP stations, but approximately half of the estimated national average dose received by the general public from exposure to natural sources. The TLDs at the Project 57 monitoring stations are exposed to both natural sources (terrestrial and cosmic) and radioactive releases from the Project 57 CA. Winds in excess of approximately 15 mph (24 km/hr) begin to generate dust movement by saltation or suspension, as determined by the monitoring instruments. Saltated particles, PM10 (i.e., inhalable dust), and PM2.5 (i.e., fine particulate dust) exhibit an approximately exponential increase with increasing wind speed. The greatest concentrations of dust occurred for winds exceeding 20 mph (32 km/hr). When the wind/dust analysis was performed for winds separated into the dominant northerly and southerly wind directions, it was evident that the northerly winds generated the higher PM10 concentrations and had the dominate influence on the PM10 concentration determined for all wind speeds when all wind directions were evaluated together. During the year, winds in excess of 20 mph (32 km/hr) occurred approximately one percent of the time. Although winds sufficient to generate dust occur at the Project 57 site, they are infrequent and of short duration. A preliminary assessment of individual wind events suggests that dust generation in response to specific wind conditions is highly variable. Wind speeds that would be expected to generate noticeable dust may not do so in every instance. Additionally, saltation transport in response to wind conditions was less common than expected. These variations were likely because of the influence of meteorological and environmental parameters other than wind.

54 ENVIRONMENTAL SCIENCES↗

ISOCS and SNAP™. Fundamemtals of nondestructive assay for international safeguards

This report introduces the principles of ISOCS and SNAP™ and their applications for measuring nuclear material. The ISOCS analysis process and the SNAP™ four-step process are described. The goals of the report are to explain 1) how ISOCS generates an efficiency curve, 2) how to calculate a minimum detectable activity (MDA) in SNAP™ and ISOCS, and 3) the corrections necessary to perform a simple assay estimate by hand calculation.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

sUAS Support for Mars 2020

The release of airborne radioactive material presents a health risk hazard to many individuals, including emergency responders and the public. Characterizing the unknown radioactive dangers produced during such incidents is essential. An advantage to utilizing unmanned aerial systems in this effort is the reduction of risk to personnel otherwise tasked with characterization, as well as more quickly obtaining data points during cloud immersion. By pairing a continuous air monitor (CAM) and detector with a drone, quantification of unknown material in the plume can be determined, and thus the extent of dose to people in the area can be estimated. A small, low-flow air sampler with a Geiger-Mueller counter may be characterized using experimental measurements of point sources representing a cloud of material and modelled using the Monte Carlo N-Particle Transport Code (MCNP), a software package for simulating nuclear processes, to find the usefulness and advantages of the system for detection of nuclides of concern in various nuclear plume scenarios. The flying time of small-unmanned aircraft systems (sUASs) is limited; thus, the minimum detectable activity and concentration rates possible must be determined over the system’s window of operation. While the activity concentrations corresponding to external and internal doses of concern can be detected with certainty for 90 Sr and 137 Cs in a plume, those corresponding to 238 Pu are more difficult to detect. The central principle of employing the sUAS is that the cockpit does not have an operator onboard the aircraft; therefore, control of the aircraft must take place by other means. With a sUAS, command and control can be separated into three distinct methods: (1) ground control or remote piloting, (2) semi-autonomous control, and (3) autonomous control. The isotope 238 Pu is used in the space industry, providing long-lived heat generation for fuel in outer space (NASA 2014). Due to the inherent risks with rocket launches to get objects into space, monitoring for release of these hazardous substances is a major undertaking. The current means is through employment of environmental CAMs. It is conceivable, however, that drone monitoring for release of 238 Pu in the event of a space launch mishap would give a greater chance of detection of plutonium particulate, because the particulate’s diffusion in air upon settling to the ground would make the probability of collecting a particle on a filter in a CAM very low.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Neural Network Approaches for Mobile Spectroscopic Gamma-Ray Source Detection

Artificial neural networks (ANNs) for performing spectroscopic gamma-ray source identification have been previously introduced, primarily for applications in controlled laboratory settings. To understand the utility of these methods in scenarios and environments more relevant to nuclear safety and security, this work examines the use of ANNs for mobile detection, which involves highly variable gamma-ray background, low signal-to-noise ratio measurements, and low false alarm rates. Simulated data from a 2” × 4” × 16” NaI(Tl) detector are used in this work for demonstrating these concepts, and the minimum detectable activity (MDA) is used as a performance metric in assessing model performance.In addition to examining simultaneous detection and identification, binary spectral anomaly detection using autoencoders is introduced in this work, and benchmarked using detection methods based on Non-negative Matrix Factorization (NMF) and Principal Component Analysis (PCA). On average, the autoencoder provides a 12% and 23% improvement over NMF- and PCA-based detection methods, respectively. Additionally, source identification using ANNs is extended to leverage temporal dynamics by means of recurrent neural networks, and these time-dependent models outperform their time-independent counterparts by 17% for the analysis examined here. The paper concludes with a discussion on tradeoffs between the ANN-based approaches and the benchmark methods examined here.

Bilton, Kyle J. (ORCID:0000000184553689)↗

Application of spectral summing to indeterminate suspect low-level drums at Los Alamos National Laboratory

An analytical technique developed by Pajarito Scientific Corporation (PSC), utilizing spectral summing of spectra from groups of drums of similar waste type, is being employed by the Waste Disposition Project - Low Level Waste Disposal (WDP-LLWD) Group at Los Alamos National Laboratory (LANL). This technique has been used to disposition low-level radioactive waste that has dropped out of the transuranic (TRU) category and has no place to go unless it can be proven to be LLW and not TRU. The TRU program at LANL run by Mobile Characterization Services (MCS) employs two High Efficiency Neutron Counters (HENC) with built-in gamma assay systems to assay radioactive waste for shipment and disposal as TRU waste at the Waste Isolation Pilot Plant (WIPP) at Carlsbad, New Mexico. As well as being certified for WIPP assays, the HENC systems can also be used for low-level waste assays for disposal at LANL or off-site disposal facilities, such as the Nevada Test Site (NTS). Some of the waste processed through the HENC systems cannot be confinned TRU due to the absence of detected TRU alpha emitters above the TRU cutoff of 100 nCi/g. This waste becomes suspect low-level waste (SLLW). In many cases, the waste also can't be classified as LLW because the minimum detectable activity (MDA) of TRU radio nuclides is above the 100 nCi/g level. These wastes that do not have enough detectable TRU activity to be classified as TRU waste and have too high a MDA to be classified as LLW enter a radioactive waste characterization indetenninate status that prevents their dispositioning as either TRU waste or LLW. Spectral summing allows an experienced ganuna spectroscopy analyst to add the HENC gamma spectra of a number of similar waste items together to form a consolidated (summed) spectrum. This summed spectrum contains the assay results of the group of items rather than the individual item, and gamma peaks that were not discernable in the individual spectra can become quantifiable in the summed spectrum and the MDA for group sum is reduced. The group of waste items can then be properly classified as LLW based on the summed spectrum and valid assay values can be assigned for disposal. This technique has been successfully applied to a set of 52 debris drums - with individual MDA > 100 nCi/g - with a resulting group total TRU alpha activity concentration below 40nCi/g. Further application of the technique at LANL to other waste drums that are measured on a WIPP certified HENC system is planned and good candidate drum sets are being evaluated as indeterminate situations develop.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Application of spectral summing to indeterminate suspect low-level drums at Los Alamos National Laboratory

The spectral summing technique developed by Pajarito Scientific Corporation (PSC) is a unique modeling technique that is being employed by the Waste Disposition Project - Low Level Waste Disposition (WDP-LLWD) Group at Los Alamos National Laboratory (LANL). This technique has been used to disposition low-level radioactive waste that has dropped out of the transuranic (TRU) category and has no disposal path unless it can be proven to be LLW and not TRU. The TRU program at LANL run by Mobile Characterization Services (MCS) employs High Efficiency Neutron Counters (HENC) with built-in gamma assay systems to assay radioactive waste for shipment and disposal as TRU waste at the Waste Isolation Pilot Plant (WIPP) at Carlsbad, New Mexico. As well as being certified for WIPP assays, the HENC systems can also be used for low-level waste assays for disposal at LANL or off-site disposal facilities, such as the Nevada Test Site (NTS). Some of the waste processed through the HENC systems cannot be confirmed TRU due to the absence of detected TRU alpha emitters above the TRU cutoff of 100 nCi/g. This waste becomes suspect low-level waste (SLLW). In many cases, the waste also can't be classified as LLW because the minimum detectable activity (MDA) of TRU radionuclides is above the 100 nCi/g level. These wastes that do not have enough detectable TRU activity to be classified as TRU waste and have TRU MDAs > 100nCi/g enter a radioactive waste characterization indeterminate state that prevents their dispositioning as either TRU waste or LLW. Spectral summing allows an experienced gamma spectroscopy analyst to add the HENC gamma spectra of a number of similar waste items together to form a consolidated (summed) spectrum. This summed spectrum contains the assay results of the group of items rather than the individual item, and gamma peaks that were not discemable in the individual spectra become quantifiable in the summed spectrum and the MDA for the group sum is reduced. The group of waste items can then be properly classified as LLW waste on the summed spectrum and valid assay values can be assigned for disposal. This technique has been successfully applied to a set of 52 debris drums - with individual MDA > 100nCi/g - with a resulting group total TRU alpha activity concentration below 40nCi/g. Further application of the technique at LANL to other debris drums and sludge drums that were measured on a WIPP certified HENe is planned and good candidate drum sets are being evaluated.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

RAPID RADIOCHEMICAL ANALYSES IN SUPPORT OF FUKUSHIMA NUCLEAR ACCIDENT

There is an increasing need to develop faster analytical methods for emergency response, including emergency soil and air filter samples. The Savannah River National Laboratory (SRNL) performed analyses on samples received from Japan in April, 2011 as part of a U.S. Department of Energy effort to provide assistance to the government of Japan, following the nuclear event at Fukushima Daiichi, resulting from the earthquake and tsunami on March 11, 2011. Of particular concern was whether it was safe to plant rice in certain areas (prefectures) near Fukushima. The primary objectives of the sample collection, sample analysis, and data assessment teams were to evaluate personnel exposure hazards, identify the nuclear power plant radiological source term and plume deposition, and assist the government of Japan in assessing any environmental and agricultural impacts associated with the nuclear event. SRNL analyzed approximately 250 samples and reported approximately 500 analytical method determinations. Samples included soil from farmland surrounding the Fukushima reactors and air monitoring samples of national interest, including those collected at the U.S. Embassy and American military bases. Samples were analyzed for a wide range of radionuclides, including strontium-89, strontium-90, gamma-emitting radionuclides, and plutonium, uranium, americium and curium isotopes. Technical aspects of the rapid soil and air filter analyses will be described. The extent of radiostrontium contamination was a significant concern. For {sup 89,90}Sr analyses on soil samples, a rapid fusion technique using 1.5 gram soil aliquots to enable a Minimum Detectable Activity (MDA) of <1 pCi {sup 89,90} Sr /g of soil was employed. This sequential technique has been published recently by this laboratory for actinides and radiostrontium in soil and vegetation. It consists of a rapid sodium hydroxide fusion, pre-concentration steps using iron hydroxide and calcium fluoride precipitations, followed by Sr-Resin separation and gas flow proportional counting. To achieve a lower detection limit for analysis of some of the Japanese soil samples, a 10 gram aliquot of soil was taken, acid-leached and processed with similar preconcentration chemistry. The MDA using this approach was ~0.03 pCi/g (1.1 mBq/g)/, which is less than the 0.05-0.10 pCi/g {sup 90}Sr levels found in soil as a result of global fallout. The chemical yields observed for the Japanese soil samples was typically 75-80% and the laboratory control sample (LCS) and matrix spike (MS) results looked very good for this work Individual QC results were well within the ± 25% acceptable range and the average of these results does not show significant bias. Additional data for a radiostrontium in soil method for 50 gram samples will also be presented, which appears to be a significant step forward based on looking at the current literature, with higher chemical yields for even larger sample aliquots and lower MDA. Hou et al surveyed a wide range of separation methods for Pu in waters and environmental solid samples. While there are many actinide methods in the scientific literature, few would be considered rapid due to the tedious and time-consuming steps involved. For actinide analyses in soil, a new rapid method for the determination of actinide isotopes in soil samples using both alpha spectrometry and inductively-coupled plasma mass spectrometry was employed. The new rapid soil method utilizes an acid leaching method, iron/titanium hydroxide precipitation, a lanthanum fluoride soil matrix removal step, and a rapid column separation process with TEVA Resin. The large soil matrix is removed easily and rapidly using these two simple precipitations with high chemical recoveries and effective removal of interferences. Vacuum box technology and rapid flow rates were used to reduce analytical time. Challenges associated with the mineral content in the volcanic soil will be discussed. Air filter samples were reported within twenty-four (24) hours of receipt using rapid techniques published previously. The rapid reporting of high quality analytical data arranged through the U.S. Department of Energy Consequence Management Home Team was critical to allow the government of Japan to readily evaluate radiological impacts from the nuclear reactor incident to both personnel and the environment. SRNL employed unique rapid methods capability for radionuclides to support Japan that can also be applied to environmental, bioassay and waste management samples. New rapid radiochemical techniques for radionuclides in soil and other environmental matrices as well as some of the unique challenges associated with this work will be presented that can be used for application to environmental monitoring, environmental remediation, decommissioning and decontamination activities.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

A Bayesian Method for Identifying Contaminated Detectors in Low-Level Alpha Spectrometers

Analyses used for radiobioassay and other radiochemical tests are normally designed to meet specified quality objectives, such relative bias, precision, and minimum detectable activity (MDA). In the case of radiobioassay analyses for alpha emitting radionuclides, a major determiner of the process MDA is the instrument background. Alpha spectrometry detectors are often restricted to only a few counts over multi-day periods in order to meet required MDAs for nuclides such as plutonium-239 and americium-241. A detector background criterion is often set empirically based on experience, or frequentist or classical statistics are applied to the calculated background count necessary to meet a required MDA. An acceptance criterion for the detector background is set at the multiple of the estimated background standard deviation above the assumed mean that provides an acceptably small probability of observation if the mean and standard deviation estimate are correct. The major problem with this method is that the observed background counts used to estimate the mean, and thereby the standard deviation when a Poisson distribution is assumed, are often in the range of zero to three counts. At those expected count levels it is impossible to obtain a good estimate of the true mean from a single measurement. As an alternative, Bayesian statistical methods allow calculation of the expected detector background count distribution based on historical counts from new, uncontaminated detectors. This distribution can then be used to identify detectors showing an increased probability of contamination. The effect of varying the assumed range of background counts (i.e., the prior probability distribution) from new, uncontaminated detectors will be is discussed.

alpha↗