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All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Using spatial statistical methods to determine an effective minimum detectable activity - 15681

Instrument minimum detectable activity (MDA) is generally established by collecting multiple measurements over time of a single representative background location to estimate a probability distribution of instrument readings. A reading that is considered to be greater than would be expected from the background probability distribution (i.e., a low-probability event) is deemed to be sufficiently unlikely that it must be attributable to added radioactivity and is therefore used as the instrument MDA. An effective MDA can be determined from spatially distributed data that are much lower than the instrument MDA. This has been accomplished using a large number of spatially distributed data points (from a radiation survey) where patterns of elevated survey readings can be distinguished from background at levels. Given the variation of the instrument response and the relatively low regulatory action levels, it is vital to establish MDA thresholds that are well below corrective action levels. Processing of radiological survey data with the use of histogram binning and geo-statistics in enabling pattern recognition can be used to establish effective field MDAs that are much lower than instrument MDAs. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2018

Testing Minimum Detectable Activity of Whole-Body Counting System in the Presence of Interfering Factors

The recently updated technical standard for the Department of Energy Laboratory Accreditation Program (DOELAP) may soon require accredited laboratories to empirically verify the estimated minimum detectable activity (MDA) for the nuclides of interest measured by in-vivo detection systems. The Radiation Protection Sample Diagnostics (RPSD) program is the SNL on-site DOELAP accredited laboratory that provides in-vivo measurements of ingested gamma-emitting nuclides (or to prove the lack of significant ingested gamma-emitting nuclides) for the internal dosimetry program administered by Radiation Protection Dosimetry Program (RPDP). Currently, the main nuclides of concern for RPDP include cesium-137 and cobalt-60 as specified in the Statement of Work between the two programs. Historically, MDAs for the RPSD whole-body counting system (WBC) were calculated annually as a-priori values by averaging the critical levels (LC) of any twelve subjects with undetected Co-60 and Cs-137 and assuming MDA is twice the decision level. The purpose of this technical basis document is to evaluate the method and process that validates the a-priori MDA of the RPSD WBC.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2023

Evaluation of Minimum Detectable Activities for Stack Sample Analyses

The Rad-NESHAP program takes samples at major point sources to track how much of select nuclides are being emitted out of the stack. Samples are taken at the site where the radiological operations are being performed. The samples are then sent to analytical laboratories for analysis to determine what nuclides exist in the sample. However, due to the sensitivity of the instruments, false positives are often reported. To reduce the number of false positives, the results reported from the analytical labs are compared to set detection limits. Anything below the set detection limit is considered part of the background while anything above the limit is considered a positive hit. Therefore, the detection limit must be set low enough that any false negatives do not result in a significant dose missed at the end of the year but high enough that there are not enough false positives to skew the annual dose being reported. All detection limits listed in this section for the analyses below were designed to meet the stated equivalent emissions rate and subsequent off-site doses. These minimum detectable activities (MDAs) were developed in the 1990s and documented in the memo cited for each analysis below. These limits and memos are all currently referenced in the Quality Assurance Project Plan (QAPP) for the Rad-NESHAP Compliance Task, which is being updated to a Project Implementation Plan (PIP) concurrently with this document. As part of this update process, the original MDAs defined in this section needed to be verified that they were still conservative and meet the off-site dose levels stated.

2019 MDA↗
DOE OSTI2015

Minimum Detectable Activity for Tomographic Gamma Scanning System

For any radiation measurement system, it is useful to explore and establish the detection limits and a minimum detectable activity (MDA) for the radionuclides of interest, even if the system is to be used at far higher values. The MDA serves as an important figure of merit, and often a system is optimized and configured so that it can meet the MDA requirements of a measurement campaign. The non-destructive assay (NDA) systems based on gamma ray analysis are no exception and well established conventions, such the Currie method, exist for estimating the detection limits and the MDA. However, the Tomographic Gamma Scanning (TGS) technique poses some challenges for the estimation of detection limits and MDAs. The TGS combines high resolution gamma ray spectrometry (HRGS) with low spatial resolution image reconstruction techniques. In non-imaging gamma ray based NDA techniques measured counts in a full energy peak can be used to estimate the activity of a radionuclide, independently of other counting trials. However, in the case of the TGS each “view” is a full spectral grab (each a counting trial), and each scan consists of 150 spectral grabs in the transmission and emission scans per vertical layer of the item. The set of views in a complete scan are then used to solve for the radionuclide activities on a voxel by voxel basis, over 16 layers of a 10x10 voxel grid. Thus, the raw count data are not independent trials any more, but rather constitute input to a matrix solution for the emission image values at the various locations inside the item volume used in the reconstruction. So, the validity of the methods used to estimate MDA for an imaging technique such as TGS warrant a close scrutiny, because the pair-counting concept of Currie is not directly applicable. One can also raise questions as to whether the TGS, along with other image reconstruction techniques which heavily intertwine data, is a suitable method if one expects to measure samples whose activities are at or just above MDA levels. The paper examines methods used to estimate MDAs for a TGS system, and explores possible solutions that can be rigorously defended.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Minimum detectable activity for tomographic gamma scanning system - 15648

For any radiation measurement system, it is useful to explore and establish the detection limits and a minimum detectable activity (MDA) for the radionuclides of interest, even if the system is to be used at far higher values. The MDA serves as an important figure of merit, and often a system is optimized and configured so that it can meet the MDA requirements of a measurement campaign. The non-destructive assay (NDA) systems based on gamma ray analysis are no exception and well established conventions, such the Currie method, exist for estimating the detection limits and the MDA. However, the Tomographic Gamma Scanning (TGS) technique poses some challenges for the estimation of detection limits and MDAs. The TGS combines high resolution gamma ray spectrometry (HRGS) with low spatial resolution image reconstruction techniques. In non-imaging gamma ray based NDA techniques measured counts in a full energy peak can be used to estimate the activity of a radionuclide, independently of other counting trials. However, in the case of the TGS each 'view' is a full spectral grab (each a counting trial), and each scan consists of 150 spectral grabs in the transmission and emission scans per vertical layer of the item. The set of views in a complete scan are then used to solve for the radionuclide activities on a voxel by voxel basis, over 16 layers of a 10x10 voxel grid. Thus, the raw count data are not independent trials any more, but rather constitute input to a matrix solution for the emission image values at the various locations inside the item volume used in the reconstruction. So, the validity of the methods used to estimate MDA for an imaging technique such as TGS warrant a close scrutiny, because the pair-counting concept of Currie is not directly applicable. One can also raise questions as to whether the TGS, along with other image reconstruction techniques which heavily intertwine data, is a suitable method if one expects to measure samples whose activities are at or just above MDA levels. The paper examines methods used to estimate MDAs for a TGS system, and explores possible solutions that can be rigorously defended. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2018

Investigation of Archived Gamma Data to Report Minimum Detectable Activities for Gamma-Emitting Radionuclides in Groundwater Samples

Los Alamos National Laboratory (LANL) has performed gamma analyses on ground water samples from the Nevada National Security Site (NNSS) (previously known as the Nevada Test Site or NTS) for the last 50 years. During that time, LANL has reported all of the detected radionuclides in various reports, letters and publications. However, until recently, no minimum detectable activities (MDAs) for undetected radionuclides of interest have been reported. In fact, most gamma emitting radionuclides were not detected in groundwater samples above their respective MDAs, and the MDAs can be used to define upper -bound activities for these radionuclides in the groundwater samples.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

SU-E-I-80: Beta-Minus Emitting Radiotracers Improves Molecular Endoscopy

Purpose: Molecular Endoscopy using Cerenkov Luminescence can be used to monitor the distribution of many clinically-available PET and SPECT probes for endoscopic applications. A main limitation of Cerenkov is its limited sensitivity to small concentrations of radiotracer when using light guides s. Herein we demonstrate that the use of a high energy beta emitting radioisotope, exemplified here with 90Y provides superior sensitivity to 18F because of its higher light output and its lack of corresponding gamma emission. Methods: A series of phantom experiments were performed to compare the sensitivity and noise of the CLE system for imaging 90Y and 18F. Three vials of known concentrations of 90Y (0.008 μCi, 0.08 μCi, 1 μCi) were placed in centrifuge tubes and isolated from each other. One vial of 18F (100 μCi) was placed in the imaging chamber and imaged over the course of decay (19 hours, 43 minutes, or ∼10 half-lives). Image time-points were formed from 5-minute integrations. Results: Using an SNR of 10 to define the noise-floor, the 90Y minimum detectable activity was 0.056 μCi. To the contrast, the minimum detectable activity for 18F was 11.63 μCi. These data demonstrate a 207-fold improvement in SNR of 90Y compared to 18F, when controlled for activity. Conclusion: This study demonstrated that a pure β- radionuclide such as 90Y be used is superior to 18F for Cerenkov Endoscopy. Further study is needed to demonstrate its utility in preclinical studies, endoscopic applications, intraoperative, and radiotherapy applications.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2017

A review of the developments of radioxenon detectors for nuclear explosion monitoring

Developments in radioxenon monitoring since the implementation of the International Monitoring System are reviewed with emphasis on the most current technologies to improve detector sensitivity and resolution. The nuclear detectors reviewed include combinations of plastic and NaI(Tl) detectors, high purity germanium detectors, silicon detectors, and phoswich detectors. The minimum detectable activity and calibration methods for the various detectors are also discussed.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2019

Tank 22 Supernate Sample Characterization for Select Radionuclides

The Savannah River National Laboratory (SRNL) was requested by Savannah River Remediation (SRR), through a Technical Assistance Request (G-TAR-H-00007) to provide additional sample characterization on unfiltered Tank 22 supernate identified as sample HTF-22-18-117, which was delivered to SRNL Shielded Cells on December 14, 2018. A summary of the average analytical results for this unfiltered Tank 22 supernate liquid follows. The density of the unfiltered Tank 22 supernate is 1.008 g/mL, 0.2%RSD. The Tank 22 supernate sample was analyzed for a total of 45 radionuclide species, with 15 of these radionuclides showing activity levels above minimum detectable activity of at least one replicate. Cesium-137 showed the highest activity at 2.90E+07 dpm/mL, 2.35 %RSD (1.30E-02 Ci/ L) with U-235 showing the lowest measurable activity at an average of 1.66E-01 dpm/mL, 6.26 %RSD (7.49E-11 Ci/L). The primary beta emitting radionuclides in the unfiltered Tank 22 supernate sample include Sr-90, Y-90 and Cs-137 at average activity concentrations of 1.10E-05, 1.10E-05 and 1.30E-02 Ci/L, respectively. The primary gamma emitting radionuclide is Ba-137 m at an average activity concentration of 1.23E-02 Ci/L (note that the activity concentration for Ba-137 m was calculated as 94.7% the Cs-137 activity concentration).

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Activation product analysis in a mixed sample containing both fission and neutron activation products

Activation analysis of gold (Au) is used to estimate neutron fluence resulting from a criticality event; however, such analyses are complicated by simultaneous production of other gamma-emitting fission products. Confidence in neutron fluence estimates can be increased by quantifying additional activation products such as platinum (Pt), tantalum (Ta), and tungsten (W). This work describes a radiochemical separation procedure for the determination of these activation products. Anion exchange chromatography is used to separate anionic forms of these metals in a nitric acid matrix; thiourea is used to isolate the Au and Pt fraction, followed by removal of the Ta fraction using hydrogen peroxide. W, which is not retained on the first anion exchange column, is transposed to an HCl/HF matrix to enhance retention on a second anion exchange column and finally eluted using HNO3/HF. Chemical separations result in a reduction in the minimum detectable activity by a factor of 287, 207, 141, and 471 for 182Ta, 187W, 197Pt, and 198Au respectively, with greater than 90% recovery for all elements. These results represent the highest recoveries and lowest minimum detectable activities for 182Ta, 187W, 197Pt, and 198Au from mixed fission-activation product samples to date, enabling considerable refinement in the measurement uncertainties for neutron fluences in highly complex sample matrices.

anion exchange↗
DOE OSTI2018

GEANT4 investigation of 40 K and 222 Rn interference effects on actinide detection sensitivities of CEMRC BEGe lung detectors

Internal dosimetry measurements of trans-uranic (TRU) radionuclides within the human body can be accomplished by γ-ray spectroscopy using high-purity germanium detectors. While measurements are typically conducted inside shielded chambers in order to minimize the background detector signal, the presence of radioactive 40 K within the body cannot be removed, which produces a high-energy γ ray at 1461 keV. The 40 K, in addition to airborne 222 Rn, contributes to the radiation field within the chamber that increases to the minimum detectable activity that can be achieved in practice. The objective of the present work is to estimate the contributions from 40 K internal to the human body and airborne 222 Rn to the background signal in a lung counter germanium detector array using GEANT4 simulations, with a focus on TRU isotope energy region (17–59 keV). The investigations consisted of three main simulations: 1. Simulating the γ-ray flux at the geometric center of the CEMRC counting chamber arising from a 25 nCi source of 40 K from (a) a point source, and (b) a source distributed throughout a BOMAB phantom. 2. Simulating the response of the CEMRC BEGe detectors arising from a 25 nCi source of 40 K from (a) a point source, and (b) a source distributed throughout a BOMAB phantom. 3. Simulating the response of the CEMRC BEGe detectors to the environmental background in the counting chamber, consisting of airborne 222 Rn (0–4 pCi/L) and 40 K (0–160 nCi) internal to the human body. The results are used to investigate the 40 K background contribution to a routine spectrum recorded by the lung detectors at the CEMRC Internal Dosimetry laboratory over a typical half hour counting interval used in the dosimetry of radiological workers. The minimum detectable activity was calculated for various 222 Rn and 40 K activities and compared to experimentally determined values. The contribution of cosmic rays was not included in this investigation.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2024

Evaluation of Significance Tests on Stack Sample Data

In October 2023, external assessors visiting LANL’s Rad-NESHAP team reviewed the annual source term documentation package for stack emissions. This document discusses how the team uses significance tests to determine whether or not a measurement represents an actual detection of emitted radioactive material. The significance tests used in the evaluation compare the measured sample result to the Minimum Detectable Activity (MDA) for the sample; to the uncertainty of that measurement; and to the mean blank sample for the same time period. The blank test uses the average blank plus 2-standard deviations (2σ) of the blank results as a comparison threshold for this final evaluation.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2018

Anticoincidence analysis for radioxenon detection

A number of beta-gamma coincidence detectors have been deployed for monitoring nuclear explosions. When 133Xe is present in a sample it causes interference in the calculation of the metastable isotopes. Therefore, an alternative method for activity quantification is to analyze the anticoincidence spectrum of the plastic detector to decrease interference from 133Xe. Simulated and experimental measurements were tested. The results show that the anticoincidence analysis produces more accurate results when quantifying the activity of 131mXe but is sometimes limited by the background increasing the minimum detectable activity.

Sivels, Ciara B.↗
DOE OSTI2016

WIPP Certification of the Retrieval Box Assay System at the Advanced Mixed Waste Treatment Plant, Idaho - 16241

The Retrieval Box Assay System (RBAS) at the Advanced Mixed Waste Treatment Plant (AMWTP) is an Imaging Passive Active Neutron system supplemented with a High Purity Germanium (HPGe) gamma system for the quantification of Pu-239 Fissile Gram Equivalent (FGE) and transuranic (TRU) alpha activity of waste containers of various sizes up to 249 cm x 149 cm x 198 cm (Length x Width x Height). The system was originally calibrated in 2002 (with a range of surrogate matrices and radioactive sources in a 213 cm x 122 cm x 122 cm FRP box) and has been operational for over a decade. As of August 2015, the RBAS has measured approximately 18,000 boxes (50,000 m3 of waste) which has subsequently been fed into the facility's treatment process. Its success, and the wide envelope of waste that it can characterize, has led AMWTP to extend the RBAS mission to include WIPP certification. As part of this new project, the instrument must be capable of meeting WIPP's Waste Acceptance Criteria (WAC) which include reporting activities and masses of the 10 WIPP-tracked radionuclides. The system will be required to assay sludge and debris waste streams, shredded waste and large metal items. The container size includes Standard Waste Box (SWBs), Fiberglass Reinforced Plywood (FRPs) and large Shredder Boxes. Sludge matrices will be loaded into 210-liter (55-gallon) drums and placed inside an SWB. After assay, boxes will be over-packed into SLB-II's for shipment to WIPP. The target waste streams will contain mixtures of Weapons Grade plutonium, Heat Source plutonium (high in Pu-238 content) and Mixed Oxide (MOX). The activity of the waste is expected to encompass a wide dynamic range of activity from low-level waste (LLW) to TRU with up to several hundred grams of plutonium. Some packages will have high gamma activity associated with fission and activation products such as Cs-137 and Co-60. The existing RBAS hardware (detectors, electronics, assay chamber etc.) was retained for this project. Two lead filters were provided and installed on the existing filter wheels to attenuate the gamma ray flux from the highest activity boxes to avoid overload of the HPGe detectors. The new software suite includes PC-FRAM (Version 5), which is used for the isotopic analysis of the gamma spectra. Integration of FRAM and neutron results is performed using a software package known as RBAS WIPP Analysis Reporting Software (RWARS). This package also performs matrix dependent background adjustments. Other upgrades provided include automated tests that identify when assay parameters fall outside a predetermined range and a new Quality Checking (QC) application. Minimum Detectable Activity (MDA) values were determined for the active and passive neutron modes as a function of measured matrix correction indices. The resulting Minimum Detectable Concentration (MDC) has been demonstrated to be less than 3700 Bq/g alpha activity (100 nCi/g). Modifications were performed to update the system TMU parameters to accommodate assay of non-standard size boxes and a wider range of matrix materials. A series of calibration confirmation, measurements were performed on site in 2014 using a representative range of surrogate boxes and radioactive sources including Pu. The WIPP WAC performance criteria were successfully demonstrated up to 157 g Pu and the system passed its first Performance Demonstration Program (PDP) cycle in the Spring of 2015. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Point Source Likelihood Framework (psl) v1.0

PSL is a software framework for localizing radiation point sources in a 3D environment using a freely moving radiation detector system. The framework assumes a single point source of radiation is in the environment and uses known detector positions and orientations along with their radiation measurements to find the most likely location of the point source. This approach to the problem enables analyses that improve upon the capability of a standard Maximum Likelihood Expectation Maximization (MLEM) approach, such as setting a source detection threshold, finding a likelihood interval for the position of the source, finding a likelihood interval for the activity (intensity) of the source, determining whether more than one point source may be present, and generating a map of Minimum Detectable Activity (MDA) for a given source type. Applications include radiological and nuclear security, consequence management, decontamination and decommissioning of nuclear facilities, and radiological emergency response.

Joshi, Tenzing↗
DOE OSTI2015

Novel HPGe Probe solution for Harsh Environments

In situ measurement is a privileged way of monitoring radioactive contamination compared to analyzing samples in a distant, specialized laboratory. Scintillators based spectrometers offer small footprints and are easy to easy to use, however they do not provide an accurate nuclide identification capability and activities measurement because notably of their limited energy resolution, for instance when low minimum detectable activity (MDA) are required, or in complex mixture of sources. On the other hand, High Purity Germanium (HPGe) detectors provide unmatched nuclide identification capability with the lowest MDA but they are not always of practical use on the field because the crystal needs to be cooled down to liquid nitrogen temperature, increasing the overall weight, bulkiness and complexity of the measurement. This paper presents the configuration and performance of a novel turnkey and compact HPGe solution developed by Canberra for radionuclide identification under harsh environments. Radio-contaminations surveys now can be undertaken outdoor under various weather conditions, in contaminated areas, underground or underwater locations (including under sea water), with fast on site deployment. The spectrometer is also designed in a small diameter tubular shape to offer minimal footprint for an operation in narrow and confined spaces. Besides, this innovative design does not mitigate the performances nor the reliability experienced with standard laboratory-grade HPGe spectrometers. This achievement relies on advanced technologies such as the encapsulation of the crystal in ultra-high vacuum (UHV) which provides higher robustness and does not requires thermal cycles faced with regular HPGe equipment. It also relies on a low vibration, low consumption electrical cooler so that no liquid nitrogen is being used. The detector is connected to a state-of-the-art digital spectroscopy suite embedded in an autonomous acquisition station monitoring the cryo-cooler and including the spectroscopy software. This system presented is therefore a new solution to perform high resolution spectroscopy where no current products are compact or robust enough to be installed, for instance for soil and water decontamination measurement when anthropogenic radioactive sources must be identified from natural ones or when low contamination levels are expected, the MDA being 3 to 5 times higher than scintillator-based detectors of similar sizes. Environmental inspection such as well logging for geophysics and mining, water contamination monitoring, or decommissioning works can thus benefits from both ruggedness, mobility without compromise on the nuclide identification or activity measurement capability. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2024

Results for the April 2024 Semiannual Salt Waste Processing Facility Decontaminated Salt Solution Sample

In this Technical Report, the chemical and radionuclide contaminant results from the April 2024 Semiannual sample of the Salt Waste Processing Facility (SWPF) Decontaminated Salt Solution (DSS) salt solution are presented in tabulated form. The information from this characterization will be used by Savannah River Mission Completion (SRMC) for the transfer of aqueous waste from SWPF to the Saltstone Production Facility (SPF) where the waste will be treated and disposed in the Saltstone Disposal Facility. This Technical Report compares results, where applicable, to SPF Waste Acceptance Criteria (WAC) LIMITS and TARGETS that were established at the time the SWPF DSS sample was obtained.1 The April 2024 Semiannual sample of the SWPF DSS is a composite from the six months of SWPF processing during the First Quarter Fiscal Year 2024 (1QFY2024) and the Second Quarter Fiscal Year 2024 (2QFY2024). The following facts pertaining to the WAC are drawn from the analytical results provided in this report. WAC TARGETS and LIMITS were met for all analyzed chemical and radioactive contaminants for which the detection limits are below the WAC TARGETS and LIMITS. Nitrosamines were not detected in the SWPF DSS salt solution sample above the instrument detection limits of <1 mg/L. The minimum detection limit (<3.33E-01 pCi/mL) is reported for 94 Nb as determined from the minimum detectable activity associated with the radiochemical method used for this radionuclide. The reported detection limit is above the requested SRMC target minimum detection limit concentration. However, the minimum detection limit reported for the April 2024 semiannual SWPF DSS sample for 94 Nb is lower than the estimated detection limit of 4.38E-01 pCi/mL initially established by SRNL in 2009. Thus, per guidance from SRMC, 2 SRNL continues to achieve as low as practical detection limits for this radionuclide.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗