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All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

NWAL Validation Report for U Assay by Potentiometric Titration (Modified Davies and Gray Technique) and U and PU Assay by Isotope Dilution Mass Spectrometry

SRNL Sensing & Metrology, previously Analytical Laboratory, is a member of the IAEA Network of Analytical Laboratories for Nuclear Material Measurement that is qualified to provide Uranium and Plutonium isotopic analysis of nuclear materials by Thermal Ionization Mass Spectrometry. As part of the membership, SRNL participated in the 2019 Nuclear Material Round Robin to maintain the qualification to provide isotopic analysis. SRNL also participated with the additional intent to qualify Uranium Assay by Potentiometric Titration (Davies and Gray method) and Uranium and Plutonium Isotope Dilution by Thermal Ionization Mass Spectrometry as part of the International Atomic Energy Agency’s Network of Analytical Laboratories. The proficiency test exercise was conducted by the Office of Safeguards Analytical Services, Department of Safeguards.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Radiation Simulation using Wi-Fi Signal Strength and Frequency

For routine radiological operations, lost radioactive sources, decommissioning and cleanup of nuclear facilities, and radiological terrorist events such as Radioactive Dispersive Devices (RDDs) or Radiological Exposure Devices (REDs), individuals are tasked with accurately locating both point and distributed sources of radiation. Radiation workers must often practice these tasks, whether part of initial training or periodically as part of formal or informal exercises or drills. In such circumstances, it is desirable to have an accurate recreation of a radioactive source, including a method of detecting it, which does not include dangerous amounts of radiation. The goal of this research was to develop a radiation simulation technology that uses IEEE 802.11x W-LAN protocol, i.e. Wi-Fi sources and detectors for these, as an inexpensive, precise and accurate method of replicating sources of gamma radiation. For this work, existing 802.11x sources were used. Software was developed for use with an Android smartphone to detect the presence of these sources. The measurements followed the expected behavior as a function of distance from the sources, mimicking that of gamma radiation, thus demonstrating the fundamental principle of operation. The ultimate system will be used in an outreach program to interest students in the radiation protection field.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2020

A Rapid Sequential Separation Method for Determination of Actinides and TENORM in Fracking Wastes - 20223

With recent advances in unconventional drilling technology in the US and around the world, the risks of environmental contamination and exposure due to technologically enhanced naturally occurring radioactive materials (TENORM) from fracking wastes have also surfaced. Although the Permian Basin (southeastern New Mexico, western Texas) has long been a modest producer of oil, the advent of fracking technology a decade ago turned it into hot property for producers. A 2018 assessment of undiscovered, technically recoverable continuous oil and gas resources by the U.S. Geological Survey estimates an average 46.3 billion barrels of oil and 281 trillion cubic feet of gas. This is likely to increase the use of unconventional drilling and TENORM generation. TENORM concentrations in oil and gas exploration and production waste can be several hundred to several thousand times higher than background TENORM concentrations. Both the thorium and uranium decay chains contribute to airborne radionuclides arising from the radon gas escaping the ground and subsequently decaying as airborne particulates. Additionally, radon decay products {sup 210}Pb and {sup 210}Po can build up in scale on the internal surfaces of oil and gas handling pipes and in sludge in refineries, becoming potential inhalation and ingestion hazards for workers. Furthermore, southeastern NM is also home to world's only licensed and operating transuranic nuclear waste repository, the DoE's Waste Isolation Pilot Plant (WIPP). Plutonium isotopes ({sup 239+240}Pu) and {sup 241}Am, are expected to account for more than 99% of the total radioactivity scheduled for disposal in the WIPP repository. Thus, an improved understanding of the environmental fate and transport of TENORM, liberated by unconventional drilling, is essential to assess how best to protect individuals and the environment. In this context, accurate measurement of TENORM and actinides in environmental and biological samples is essential. In this presentation, a new sequential method for the separation and pre-concentration of actinides (Pu, Am, Np) and TENORM (Po, U and Th) derived from oil and gas exploration is proposed. The TEVA method involves a rapid co-precipitation step to remove matrix interferences from the samples, followed by plutonium oxidation state adjustment to Pu (IV) and an incubation period of ∼ 1 hour at 50-60 deg. C to allow the resultant Po (II) to oxidize into Po(IV). The polonium, neptunium, thorium and plutonium are then separated on a TEVA column, while americium and uranium are separated on a TRU column. After separation, the alpha counting source was prepared by micro-precipitation with copper sulfide (CuS) for polonium and neodymium fluoride (NdF{sub 3}) for actinides. The efficiency and reliability of the procedures were tested by analyzing filter, drinking water and frack sand samples. The method is simple, robust and can be performed quickly with excellent removal of interferences, high chemical recovery and very good alpha peak resolution. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2022

Origin of Plutonium-244 in the Early Solar System

We investigate the origin in the early Solar System of the short-lived radionuclide 244 Pu (with a half life of 80 Myr) produced by the rapid (r) neutron-capture process. We consider two large sets of r-process nucleosynthesis models and analyse if the origin of 244 Pu in the ESS is consistent with that of the other r and slow (s) neutron-capture process radioactive nuclei. Uncertainties on the r-process models come from both the nuclear physics input and the astrophysical site. The former strongly affects the ratios of isotopes of close mass ( 129 I/ 127 I, 244 Pu/ 238 U, and 247 Pu/ 235 U). The 129 I/ 247 Cm ratio, instead, which involves isotopes of a very different mass, is much more variable than those listed above and is more affected by the physics of the astrophysical site. We consider possible scenarios for the evolution of the abundances of these radioactive nuclei in the galactic interstellar medium and verify under which scenarios and conditions solutions can be found for the origin of 244 Pu that are consistent with the origin of the other isotopes. Solutions are generally found for all the possible different regimes controlled by the interval (δ) between additions from the source to the parcel of interstellar medium gas that ended up in the Solar System, relative to decay timescales. If r-process ejecta in interstellar medium are mixed within a relatively small area (leading to a long δ), we derive that the last event that explains the 129 I and 247 Cm abundances in the early Solar System can also account for the abundance of 244 Pu. Due to its longer half life, however, 244 Pu may have originated from a few events instead of one only. If r-process ejecta in interstellar medium are mixed within a relatively large area (leading to a short δ), we derive that the time elapsed from the formation of the molecular cloud to the formation of the Sun was 9-16 Myr.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2020

The Cost of (In-) Accurate Waste Characterization in the U.S. Nuclear Power Industry - 20066

Proper disposal of waste has never been a human priority. Our history is full of examples of throwing away things with minimal effort or thought to consequence. That is until disease or toxicity become apparent and we realize we need to change. Unfortunately, we almost never have all the information we need at the time we make decisions. In addition to the technical and scientific issues, profits and politics are also factors. Newton's third law also applies to politics and so for every policy (action) put in place, there will be (opposite reaction) forces in play to push back against it. These forces are seldom balanced and so the reaction to the discovery of harm is to over-correct and destroy the benefit to be had. This of course causes those who want the benefit, or profit from it, to work to undermine or repeal those measures put in place to eliminate the harm. The issue is exacerbated when the problem to be solved is highly technical and at the edge of our science so that the harm and benefit cannot be easily or accurately quantified. At these times, reason often goes out the window and our decision makers are influenced by either greed or fear as these emotions are much easier to stoke. Those of us in the middle, with the charge to 'do the right thing for the least cost', are therefore frequently constrained and forced into actions that are neither technically the right thing to do and not cost effective. The cost of radioactive waste disposal is based on the concentrations of various radioisotopes in the waste. Some of these isotopes are relatively easy to quantify and some are not. Most laboratory methods for quantification are limited in the amount of activity that can be present during the measurement process. Current nuclear plants were not designed with the idea that taking truly representative waste samples would be important. The limits used to define waste class and therefore cost are precisely defined but the methods for determining those concentrations are far less precise. Improving the accuracy of radioactivity concentrations in waste and having consistent and reasonable oversight can reduce waste costs through active management of the process. So what can we do about it? Science may provide an answer. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Abundances of Uranium and Thorium Elements in Earth Estimated by Geoneutrino Spectroscopy

Abstract The decay of the primordial isotopes 238 U, 235 U, 232 Th, and 40 K has contributed to the terrestrial heat budget throughout the Earth's history. Hence, the individual abundance of those isotopes are key parameters in reconstructing contemporary Earth models. The geoneutrinos produced by the radioactive decays of uranium and thorium have been observed with the Kamioka Liquid‐Scintillator Antineutrino Detector (KamLAND). Those measurements have been improved with more than 18‐year observation time, and improvement in detector background levels mainly with an 8‐year nearly reactor‐free period, which now permit spectroscopy with geoneutrinos. Our results yield the first constraint on both uranium and thorium heat contributions. The KamLAND result is consistent with geochemical estimations based on elemental abundances of chondritic meteorites and mantle peridotites. The High‐Q model is disfavored at 99.76% C.L. and a fully radiogenic model is excluded at 5.2 σ assuming a homogeneous heat producing element distribution in the mantle.

58 GEOSCIENCES↗
DOE OSTI2022

Radon-222 Charcoal Canister Steady State Model Calibrations Performed in a Highly Controlled Environmental Chamber and a Natural Indoor Environment

Charcoal canisters are a common method of 222 Rn screening. The calibrations of different batches of activated charcoal used in different canister designs are typically performed in large volume controlled environmental chambers with known and controlled radon concentration, temperature, and humidity. Radon screening could be facilitated in locations without ready access to environmental chambers if canisters could be accurately calibrated and undergo quality control in less controlled environments. This study compares charcoal canister calibrations from a highly controlled radon chamber with calibrations from a basement storage area experiencing temporally varying radon. In addition, the impacts of exposure time and spectral region of interest selection on calibration accuracy are investigated by comparing calibrations calculated using three different choices. Approximately 30 mo after calibrations were completed, groups of canisters were exposed for different durations of time to investigate measurement accuracy and calibration validity over an extended time. A digital simulation of charcoal canister kinetics was also performed to establish limits on how stable radon must be in a space for equilibrium-based calibration to be performed there. Overall, the accuracy of measurements using calibrations from each space differed by less than 10% after 2 d exposure time, showing that carefully controlled conditions are not necessary for the accurate calibration of charcoal canisters. Furthermore, measurement accuracy differed by less than 2% for different spectral region selections. Accuracy improved slightly with canister exposure duration. Simulations suggest that radon instability is most tolerable near the beginning of canister calibration exposures, but this merits further experimental study.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2024

rf chopper for prebunched radioactive ion beams

An rf chopper system is being designed for the ReAccelerator (ReA) linac at the Facility for Rare Isotope Beams at Michigan State University. The 80.5 MHz ReA radio-frequency quadrupole accelerates prebunched 16.1 MHz beams, producing four satellite bunches for every main bunch. The chopper system includes an rf deflector that kicks every bunch vertically to spatially separate main and satellite bunches. A constant magnetic field superimposed with the chopper electric field biases the beam trajectory to ensure the high-intensity bunches do not experience a net deflection and are injected straight to the ReA6 cryomodule or sent for experiments. The kicked bunches are low in intensity and will be sent to a beam dump, resulting in a clean 16.1 MHz beam structure, which allows for a reliable time-of-flight separation of the isotopes. Published by the American Physical Society 2024

Physics↗
DOE OSTI2021

SR19021 Tritium Aging of Regenerated LANA.75 (Final Project Report)

The Savannah River Tritium Enterprise (SRTE) has used the metal hydride LaNi 4.25 Al 0.75 (LANA.75) in the Tritium Facilities for over two decades. LANA.75 beds store significant quantities of tritium but have a limited service life due to the radiolytic decay of tritium to He-3 within the metal matrix. It has been shown that heating tritium-aged LANA.75 under vacuum can reverse tritium aging effects, eliminating the heel of trapped hydrogen, and restoring the reversible capacity. Additional investigation is needed to ensure there are no unexpected changes to the hydride before this restoration technique is employed in full scale beds in the Tritium Facilities. This project was to be comprised of three distinct scopes: obtain tritium aging data on the sample regenerated in 2018, regenerate a tritium-aged sample at 600 °C, and perform thermal stability testing on a non-tritiated sample. Isotherms were collected on the previously regenerated sample after approximately two years of tritium aging. Isotherms were collected at 80, 100, and 120 °C. As expected, there was a decrease in the plateau pressure, an increase in plateau slope, and a portion of the “heel” of tritium trapped in the metal had been reestablished. Unexpectedly, it appeared that the plateau had shortened at the higher tritium to metal ratios as well. This is typically seen in older samples. The second scope, to regenerate a second tritium-aged LANA.75 sample, was not completed. A Task Technical and Quality Assurance Plan was written and approved, a high temperature test cell was fabricated, several pre-job briefs were held, and the hydride sample was passivated with air. Despite these successes, the hydride sample was not recovered from the legacy test cell. The third scope was to perform thermal stability testing on a “cold” LANA.75 sample. A non-tritiated sample of LANA.75 was held at 750 °C under vacuum for 200 hours to simulate exposure to multiple regeneration evolutions. Hydride isotherm performance, chemical composition, crystallinity, particle size, and morphology are compared between the pre- and post-regeneration samples. No significant changes were observed in composition, crystallinity, or particle size. Comparison of before and after isotherms showed that performance improved rather than deteriorated during the evolution. Scanning Electron Microscopy (SEM) analysis showed small growths on the particle surface after exposure to regeneration conditions. Additional testing will be required to determine the cause of these growths.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2021

The Science of the Bomb

Uranium is a naturally occurring element with wo isotopes of interest – U238 and U235; U235 is useful for its ability to sustain a chain reaction; Chain reactions are the fundamental way that nuclear reactors work, as well as how bombs work; Enrichment of U235 is important for both nuclear energy as well as nuclear weapons; A safety focus is at the center of all nuclear operations in the United States

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2021

Recommended Updates to Solubility Controls for Modeling Leaching of Technetium, Uranium, Neptunium, Plutonium, and Iodine from the Residual Waste Layer of Closed Savannah River Site High Level Waste Tanks

The objective of this report is to relate waste leaching experiments by the Savannah River National Laboratory (SRNL) to performance assessment modeling of release of several radionuclides from residual waste in closed high-level waste tanks. SRNL leached residual waste from Tanks 18F and 12H with solutions simulating the modeled evolution of pore fluids in residual waste layers of closed tanks through various stages of grout degradation. Tc-99, uranium isotopes, Np-237, plutonium isotopes, and I-129 were analyzed in the leachate multiple times over 90 days.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2022

Solvent Recovery and Management at the Savannah River Site H-Canyon Facility

NIOWAVE, Inc. is a domestic supplier of medical and industrial isotopes from uranium and radium. The Savannah River National laboratory (SRNL) is currently providing support to NIOWAVE, which plans to deploy a superconducting electron linear accelerator (LINAC) to fission uranium for Mo-99 production without the need for a nuclear reactor or HEU. The uranium from the Mo-99 production targets will be purified using a modified PUREX (Plutonium Uranium Reduction Extraction) solvent extraction process to recover the uranium in the product stream. The uranium will then be precipitated as an oxalate which is calcined to U 3 O 8 to fabricate pellets for new Mo-99 targets. In previous support provided to NIOWAVE, the SRNL demonstrated a solvent washing process to remove degradation products from the tributyl phosphate (TBP) solvent used in the modified PUREX process under development for uranium recovery. To supplement this technology demonstration, NIOWAVE requested the SRNL to provide summary information on the solvent recovery and management activities which are used at the Savannah River Site (SRS) H-Canyon facility. An existing reference document for the reprocessing of irradiated HEU fuels at the SRS was used as the primary reference for the solvent management activities; although, other reference documents were used to provide supplementary information. The information provided includes a brief summary of the solvent degradation issues which have been observed in the H-Canyon solvent extraction cycles and resulting process safety concerns. The solvent recovery processes for the three cycles of solvent extraction used in the H-Canyon were subsequently described including the process equipment which consists of the continuous and batch solvent washers, pumps, and tanks. A final section is provided on the monitoring and analysis of solvent quality based on the previous work performed at the SRNL for NIOWAVE and past research and development activities performed to support the solvent extraction processes in both the SRS F-Canyon and H-Canyon facilities.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2022

Cell G 7606A Mockup Measurement Analysis Methods and Results

The production of research isotopes at the Radiochemical Engineering Development Center (REDC) invariably leads to the accumulation of radioactive waste. Storage of this waste onsite at ORNL serves as an interim solution prior to shipment to long term storage facilities, however onsite capacity is limited. Proper characterization of the activity of waste products is essential for determining the appropriate waste stream and, ultimately, mitigating the cost of disposal. This is typically done via gamma spectrometry and use of the In-Situ Object Counting System (ISOCS), a software package from Mirion that serves as an accepted community standard. However, variability in the contents, density, and activity distribution in waste containers can introduce large errors in ISOCS quantification. A measurement campaign using a mockup of the proposed setup in Cell G seeks to quantify the magnitude and source of these systematic errors. Analysis of this data will guide the creation of ISOCS geometry templates and measurement methods designed to minimize overall uncertainty in reported activities.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2026

Evaluation of the Radiological Characterization for Off-Site Source Recovery Program Waste Streams LA-OS-00-01.001, LA-OS-00-03, and LA-OS-00-04

The purpose of this memorandum is to satisfy the requirements of Section 4.4 of the CCP Acceptable Knowledge Documentation procedure CCP-TP-005 (Ref. 1). This evaluation is updating the previously issued memo NEN3:24-045 issued in August 2024. CCP-TP-005 requires an AK Expert and the OSRP group to evaluate the radionuclide characterization of a waste stream and prepare the NDA Memorandum (letter to CCP Records). This NDA memorandum was written with input from the OSRP group, as required by CCP-TP-005. This memo includes a discussion of the limitations for the radiological characterization and a description of the characterization method.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2020

Isotope separation methods and systems

Methods and systems for the separation of isotopes from an aqueous stream are described as can be utilized in one embodiment to remove and recover tritium from contaminated water. Methods include counter-current flow of an aqueous stream on either side of a separation membrane. The separation membrane includes an isotope selective layer (e.g., graphene) and an ion conductive supporting layer (e.g., Nafion®). An electronic driving force encourages passage of isotopes selectively across the membrane to enrich the flow in the isotopes.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2020

Determining Cs-137 Background body burdens for Wild Pigs At Savannah River Site

Savannah River Site (SRS) is a unique U.S. Department of Energy facility in that most of the 310-square miles of land has not be industrialized since operations began in 1953. This has allowed abundant, undisturbed vegetation structure to exists as well as an increase in herd sizes among wildlife populations onsite. Cs-137 is widespread across SRS due to the production of nuclear materials, but is mostly from global atmospheric fallout from nuclear weapons testing. Due to most of SRS being undisturbed land, the concentrations of Cs-137 are higher on site than off site. The SRS began conducting annual deer harvests that were open to the public beginning in 1965. All animals harvested during these hunts are monitored for Cs-137, which is measured using a Sodium Iodide detector. There has been an effective method for calculating the background Cs-137 body burden for the deer population at the SRS, developed by Gaines and Novak using data collected during hunts and gamma overflight data to determine the areas contaminated by site activities. For animals that have a Cs-137 concentration above the established background level, a dose is assigned to the hunter. By applying the same method for wild pigs using data collected from hunts, gamma overflight, and developing a site-specific home range of wild pigs, we were able to calculate the Cs-137 background body burden concentrations of wild pigs on site. Objectives: Develop site-specific home range for wild pigs at SRS; Determine the average concentrations of Cs-137 in wild pigs from background; Use Gaines and Novak method for determining background Cs-137 body burdens wild pigs on SRS. Methods and Materials: Development of site-specific home range of wild pigs was determined to be 2.5 miles using: Four studies conducted between 1969 and 1992 for the SRS wild pig population. (Mayer and Brisbin) The background compartments were determined using: Gamma overflight data showing contaminated areas; Site-specific home range of the wild pigs. Results and Conclusion: All data were generated through the JMP software to provide a histogram and summary statistics. The estimate of an upper tolerance limit for the background contamination with 95% confidence is 1.965 pCi/g for the wild pigs. This means that for wild pigs on SRS from the background compartments are expected to have Cs-137 concentrations of less than 1.965 pCi/g. The current background for the deer is 2.59 pCi/g. For the background compartments Cs-137 concentrations from 2012-2018: Mean: 0.815 pCi/g, Maximum: 2.341 pCi/g, Minimum: -0.704 pCi/g.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2020

An Experimental Feasibility Study on Applying Neutron Tomography to Encapsulated Spent Nuclear Fuel - 20050

Visual inspection makes easier to ensure the integrity and safety of spent nuclear fuel (SNF) than any classical techniques. Various classical techniques have been applied but there are no reliable methods to qualitatively and quantitatively verify spent fuel in dry storage. Thus, the present authors have developed the prototype safeguards apparatus for dry storage employing the array of He-4 gas scintillation detectors (S670E, Arktis Radiation Detectors Ltd., Switzerland), newly designed to simultaneously measure thermal and fast neutrons without any moderators. The S670E detector has a cylindrical shape with a diameter of 52 mm and active length of 600 mm (total length: 875 mm). The detector is filled by He-4 gas with an approximate pressure of 180 bar for fast neutron detection, and its inner wall is coated by Li-6 for thermal neutron detection. The scintillation lights generated via Li-6 nuclear reaction and elastic scattering are collected by 24 SiPMs linearly paired at the center of the detector. The detector delivers a TTL (Transistor-Transistor Logic) output for pulse readout and UART (Universal Asynchronous Receiver Transmitter) for device control. In order to assess feasibility of the apparatus, an experimental system has been designed, built, and optimized via computational studies. Cf-252 neutron sources and linearly arrayed detectors, working as a single detector, were occupied for this study due to the difficulties in working with SNF. The laboratory scale cask (diameter: 0.67 m, height: 1.5 m), minimized by a factor of 10 compared to the actual thickness of a commercial TN-32 cask, was also manufactured. The detector array was designed to rotate the lab-scale cask and obtain 36 image profiles at every 10 degrees. All profiles were aligned in single frame image called a sinogram, and the cross-sectional image was then fabricated by the inverse radon transform algorithm. These experiments have been repeated with different configurations and numbers of sources. Some gamma-ray sources were also measured with neutron sources in order to distinguish between neutron and gamma-ray pulses. Basically, a He-4 detector is designed to run on Linux OS so it is difficult to directly apply to Windows-based equipment widely used in S. Korea. Therefore, a new data acquisition board working on Windows OS was designed and built. The board mainly consists of FPGA (Field Programmable Gate Array) and SoC (System on Chip) for TTL pulse readout, sorting measured data, and transferring data to a user interface. In conclusion, the tomographic system at lab-scale has shown considerable potential to detect a partial or gross defect of encapsulated assemblies in dry storage. Next steps of this study will be to 1) repeatedly carry out experiments to demonstrate scientific reliability and validity, and 2) numerically integrate signals with weight factors to enhance the image quality since the suggested system based on passive interrogation method requires longer measurement time. Finally, the system will apply to a commercial dry storage phased out soon in S. Korea. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2020

Fission Product Gas Monitoring During Fuel Drying Operations - 20094

The goal of the project was to devise and implement a method of accurately qualifying any noble gas emitted during fuel drying operations (part of the dry fuel storage process) in order to provide the data needed to validate the off-site dose calculations. The imputes of the project was the NRC publishing Information Notice 18-01, 'Noble Fission Gas Releases During Spent Fuel Cask Loading Operations' in February 2018. The information notice describes some industry events that occurred during vacuum drying operations as part of a dry fuel campaign. The solution was devised by Fort Calhoun Station staff in partnership with Mirion Technologies technical staff. The monitoring was accomplished by directing all the effluent from the fuel storage cask vacuum drying machine through a sample chamber containing a compact ISOCS characterized CZT based gamma spectrometer connected to a Mirion Data Analyst. This resulted near real-time quantification of the radionuclides of concern. (authors)

07 ISOTOPE AND RADIATION SOURCES↗