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DOE OSTI2019

Risk perception and social representation on radioactive waste: a case study with students of IPEN

Proper management of radioactive waste in teaching and research laboratories involves not only techniques and procedures for the treatment, storage and disposal of these materials but also the responsibility of all those who handle them. Responsibility begins with with methods that generate less waste and with the correct segregation of contaminated materials prior to disposal, In this context, the present study aimed to know, identify, understand and analyze the perception of risk and social representations about radioactive waste and its management by students and researchers of the Nuclear and Energy Research Institute. According to Renn (1992), two of the approaches to risk perception are cultural and sociological, which served as the basis to the present study. Within this analysis, the concept of risk is understood as a phenomenon and, therefore, the phenomena experienced by the individuals are related to their life histories, social relations, religious beliefs and academic formation, elements that interfere in the perception of risk. The method used to collect data was free evocation, with participants receiving the following instructions: 'Please write five words that pop up into your mind when you listen to the words radioactive waste and, sequentially, write a sentence using these words'. All participants received instructions of the voluntary character of the participation and of the confidentiality of the data. A total of 73 students and researchers were interviewed, including scientific initiation, masters degree, doctorate and postdoctoral researchers. 353 words were mentioned, of which 209 were different from each other. After analyzing the sentences, the words were inserted in 12 categories and later analyzed in the software INVOCADO, which provides a diagram with the possible structure of the social representations of the group studied. The analysis of the four-quadrants and their categories, as well as the mentioned words that compose them, are therefore consistent with the universe of the public analyzed. While it is not the totality of students dealing with radioactive materials, in general these individuals understand the importance of technology and that the adoption of safe practices ensures the successful development of research. Future developments: Analyze quantitatively and qualitatively the other questions of the questionnaire. Analyze the data and information obtained from the participant observation, which consisted of a follow-up of three students in their laboratory practices. Propose to the institute a manual of good practices that can be used by the students in their laboratory practices.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

A radioactive tracer dilution method to determine the mass of molten salt

A new technique for molten salt mass determination, termed radioactive tracer dilution, that uses 22Na as a tracer was validated. It has been a challenging problem to determine the mass of molten salt in irregularly shaped containers, where a highly radioactive, high-temperature molten salt was used to process nuclear spent/used fuel during electrochemical recycling (pyro-processing) or for coolant/fuel salt from molten salt reactors. A radioactive source with known activity is dissolved into the salt. After a complete mixture, a small amount of the salt is sampled and measured in terms of its mass and radioactivity. By finding the ratio of the mass to radioactivity, the unknown salt mass in the original container can be precisely determined.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2026

Ultra-pure nickel for structural components of low-radioactivity instruments

The next generation of nuclear and particle physics rare-event search experiments demand structural materials combining ultra-low levels of radioactive contamination with exceptional mechanical strength. Here, this study evaluates chemical vapor deposition (CVD) nickel as a low radioactive background candidate structural material for such applications. Manufacturer-supplied CVD Ni grown on aluminum substrates was assayed via inductively coupled plasma mass spectrometry (ICP-MS) employing isotope-dilution. These material assays produced measured bulk concentration of 232 Th, 238 U, and nat K at the levels of ~70 ppq, ≲100 ppq, and ~900 ppt, respectively, which is the lowest reported in nickel. Surface-etch profiling uncovered higher concentrations of these contaminants extending ~10 μ m beneath the surface, likely associated with the aluminum growth substrate. Additionally, the CVD Ni underwent tensile testing alongside standard Ni samples. CVD Ni exhibited a planar tensile strength of ~600 MPa, significantly surpassing standard nickel. However, heat treatment was found to reduce the tensile strength to levels comparable to standard Ni, with implications for high-temperature weld joining methods. The results reported are compared to the one other well documented usage of CVD Ni in a low radioactive background physics research experiment and a discussion is provided on how the currently reported results may arise from changes in CVD fabrication or testing process. These results establish CVD Ni as a promising low-radioactivity structural material, while outlining the need for further development in surface cleaning and weld-joining techniques to fully realize its potential in large-scale, low radioactive background rare-event search experiments.

47 OTHER INSTRUMENTATION↗

DOE OSTI2015

Study on effect of geometrical configuration of radioactive source material to the radiation intensity of betavoltaic nuclear battery

Nuclear batteries have strategic applications and very high economic potential. One Important problem in application of nuclear betavoltaic battery is its low efficiency. Current efficiency of betavoltaic nuclear battery reaches only arround 2%. One aspect that can influence the efficiency of betavoltaic nuclear battery is the geometrical configuration of radioactive source. In this study we discuss the effect of geometrical configuration of radioactive source material to the radiation intensity in betavoltaic nuclear battery system. received by the detector. By obtaining the optimum configurations, the optimum usage of radioactive materials can be determined. Various geometrical configurations of radioactive source material are simulated. It is obtained that usage of radioactive source will be optimum for circular configuration.

30 DIRECT ENERGY CONVERSION↗

DOE OSTI2015

SU-E-T-534: Level of Residual Radioactivity of Activated Parts of a Decommissioned Cyclotron

Purpose: CTI cyclotron RDS-111 was used at the Hong Kong Sanatorium and Hospital (HKSH) to produce radiopharmaceuticals and radioactive tracers for diagnostic scans between 1999 and 2007. During the operation, some machine components became radioactive by activation. For the safety of staff, decommissioning took place in 2009, two years after the cyclotron had stopped operation. This study investigates the residual radioactivity and radionuclides found in different cyclotron components in 2014 in compliance with the local regulations in Hong Kong for transfer of radioactive waste. Methods: A representative sample of each part was counted using a high-purity germanium detector (manufacturer: ORTECT) for at least four hours. GammaVision, a multichannel analyzer software, was used to identify the radionuclides found in the cyclotron components, as well as the associated activities. A standard library and a Mariscotti peak search algorithm were used to identify the present radionuclides. Only radionuclides with half-life greater than 180 days were considered. Results: Among the components, the Havar target foil has the highest specific activity ((4.6±0.6)×10{sup 2} Bq/g), with Co-60 being the most prominent ((3.8±0.5)×10{sup 2} Bq/g). The total activity of the target foil, however, is still low due to its small mass of 0.04 g. Radioisotopes Mn-54 (46±6 Bq/g), Na-22 (6.8±0.8 Bq/g), Co-57 (7.3±0.9 Bq/g), and Fe-59 (6.0±0.9 Bq/g) have also been detected in the target foil. The target window holder and the vacuum window register a specific activity of 88.3±0.6 Bq/g and 48.6±0.1 Bq/g, respectively. Other components, such as the collimator, the target tube, the valve body and the beamline, are also found with trace amounts of radionuclides. Conclusion: Even seven years after the cyclotron had stopped operation, some components still exhibited residual radioactivity from activation exceeding the IAEA clearance levels. Special consideration for radiological protection may need to be taken in the disposal of the cyclotron parts.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2018

Air Dispersion Modeling of Radioactive Releases During Proposed 236Z Rubble Pile Removal Activities

This report is part of the planning process for the removal of the 236Z rubble remaining as of January 1, 2018, at the Plutonium Finishing Plant (PFP) on the Hanford Site. This report documents an analysis considering standard techniques for the removal of demolition rubble. It does not document any decisions about the removal approaches. This document is intended to guide the establishment of a boundary for the application of an air monitoring regime. Atmospheric dispersion modeling using estimated release rates has been conducted to provide information on the location and levels of radioactive contamination that may be expected as the result of rubble removal activities. Hourly meteorological data collected over a 6-year period (2004–2009) were used to examine the effects of wind speed, direction, and stability on projected concentrations of contaminants in air and deposited on nearby surfaces. The source terms modeled in this report were based on the residual transuranic contamination levels that are anticipated to remain in the 236Z rubble following building demolition. The potential airborne concentrations were established using the five-factor formula from DOE-HDBK-3010 (DOE 1994) considering material-at-risk, damage ratio, airborne release fraction, respirable fraction, and leak path factor. Radioactive contamination emissions were calculated by release mechanism and rubble pile area for on-shift and off-shift activities. The emissions from the applicable sources were combined to provide emissions estimates for each day from each cleanup area. The U.S. Environmental Protection Agency’s (EPA’s) AERMOD computer code was used to estimate atmospheric dispersion and deposition of the released radioactive materials in the immediate vicinity of the planned rubble removal activities Both airborne and surface concentrations were modeled with AERMOD. Hourly derived air concentrations (DAC) were modeled for an array of receptors covering the rubble pile site and surrounding area. Peak (95th and 99th percentile) values of time-integrated air concentrations at these receptor points are derived from these hourly values, with modeling results reported as total incremental air concentrations in DAC-hours occurring over the selected time period. Total accumulated deposition amounts were also evaluated with AERMOD using the same array of receptors, with results reported as disintegrations per minute (dpm) per 100 cm2. Each rubble pile was considered in terms of its size, location, and estimated contamination levels. The modeling was based on the assumed sequence of the rubble removal activities. The results in this report are based on cleanup schedules and methods proposed by CHPRC. As of January 1, 2018, there are broadly three large rubble piles (west, footprint, and east) remaining of the 236Z Building process cell. It is assumed that the overall cleanup will require either 1) 30 days over a 6-week period, or 2) 60 days over a 12-week period. Work was assumed to be performed during 10-hour day shifts. The analysis shows that some releases of radioactive material are to be anticipated during the cleanup of the 236Z rubble piles. The modeling results presented here are closely tied to the details of how the rubble removal is to be conducted. The results indicate that for the rubble removal, the radiological exposures from the planned activities are predicted to be well below the designated limits for air and soil exposures. The cleanup of the 236Z main process cell rubble is expected to release some alpha-emitting radionuclides. However, results based on 6 years of hourly climatological data, two representative cleanup periods, realistic mitigating actions, and a bounding source term indicate that concentrations at the fence line of the PFP facilities should remain well below 12 DAC-hours/week. The results presented in this report are a reevaluation of a subset of the results presented in PNNL-20173 Rev. 4.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Characterizations of the radioactive waste by the remotely-controlled collimated spectrometric system

Decontamination and decommissioning of the research reactors MR (Testing Reactor) and RFT (Reactor of Physics and Technology) has recently been initiated in the National Research Center (NRC) 'Kurchatov institute', Moscow. In the building, neighboring to the reactor, the storage of HLRW is located. The storage is made of monolithic concrete in which steel cells depth 4 m are located. In cells of storage the HLRW packed into cases are placed. These the radioactive waste are also subject to export on long storage in the specialized organization. For characterization of the radioactive waste in cases the remote-controlled collimated spectrometer system was used. The system consists of a spectrometric collimated gamma-ray detector, a color video camera and a control unit, mounted on a rotator, which are mounted on a tripod with the host computer. For determination of specific activity of radionuclides in cases, it is developed programs of calculation of coefficients of proportionality of specific activity to the corresponding speeds of the account in peaks of full absorption at single specific activity of radionuclides in cases. For determination of these coefficients the mathematical model of spectrometer system based on the Monte-Carlo method was used. Dependences of calibration coefficients for various radionuclides from distance between the detector and a case at various values of the radioactive waste density in cases are given. Measurements of specific activity in cases are taken and are discussed. By results of measurements decisions on the appeal of the radioactive waste being in cases are made. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

ALARA on the radioactive waste management activities at NPP in Korea - 15189

This study examined and analyzed on-site experience to investigate the effect of the 'as low as reasonably achievable' (ALARA) principle on the level of radiation exposure management during NPPs' radioactive waste management activities. In particular, the ALARA principle was observed by various activities of radioactive waste management. According to 2012 data, workers' radiation exposure due to radioactive waste management activities in one light water reactor (LWR) unit amounted to a collective dose of 14.8 man-mSv, and this value corresponded to approximately 3.5% of the radiation exposure due to the NPP's total activities. The present study is expected to contribute to the reduction of workers' radiation exposure by analyzing and synthesizing field experience from the perspective of the ALARA principle during radioactive waste management, and transmitting the results to other NPPs. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Design and performance testing of a novel far field Gamma system to assay radioactive waste in 400 and 200 litre drums - 15339

A Far Field Gamma ray Measurement system has been developed for measuring and assaying both 400 and 200 litre waste drums with a wide range of activity at nuclear power stations. These include radioactive waste drums with surface dose-rates of less than 2 mSv/h (200 mrem/h) and, at the higher range of activity, 400 litre drums with surface dose-rates of up to 50 mSv/h (5 rem/h). The assay system incorporates an electro-mechanically cooled high purity Germanium profile detector and advanced digital spectroscopic analysis electronics. A novel optimised conical-trapezoidal collimator has been incorporated to reduce the effect of the radioactive background in addition to the conventional detector lead shield with tin and copper graded lining to reduce the effect of lead X-rays. Two features have been incorporated to reduce detector dead time and maintain detector resolution when measuring high activity drums. The first is the inclusion of two automated tungsten filters of different thickness as an integral part of the detector collimator. The filters are used in conjunction with an automated rail system on which the detector platform is mounted. The detector can be positioned at different distances from the drum surface in order to reduce detector dead time. The deployment of the filters and the adjustment of the drum - detector position are automatically controlled based on a user defined dose-rate limit table. The data is provided by two Geiger-Muller dose-rate detectors, which are used to measure the drum surface dose-rate. This data is subsequently used to adjust the detector position and in cases of higher dose-rate deploy one or other of the tungsten filters. The drum rotation platform incorporates a load cell to determine drum weight and hence the drum density. This information is used to calculate an attenuation correction based on the assumption of uniform drum density and uniform distribution of activity within the drum. Where the waste drums have a more complicated and known regular internal structure, such as a small radioactive region surrounded by an annulus of shielding material, the analysis algorithms incorporated into the far field geometry and attenuation correction spectroscopic data analysis code are able to make a more accurate determination of drum activity than would be possible assuming a completely uniform drum. The detector energy and efficiency calibration is achieved using a point source with multiple gamma ray energy peaks. The response of the high purity Germanium detector is also modelled using the Monte Carlo Neutron Photon code and the model is benchmarked using the measured point source calibration data. The performance of the instrument has been determined for a range of uniform density matrix drums containing a set of Eu-152 line or rod sources located in re-entrant tubes positioned on an equal volume basis. When rotated, the test drums with volume distributed line sources simulate waste drums of different density each with a uniformly distributed radioactive source. The measurements have been validated using Monte Carlo simulations. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

The Approach of the Environmental Consequences and Impacts during Transport of Radioactive Materials (RAM)-A Safety Case - 16420

International Atomic Energy Agency (IAEA) works, together with Member States (MS), to provide a strong, sustainable and visible global nuclear safety and security framework, protecting people and the environment from the harmful effects of ionizing radiation. In the announcement of this IAEA Conference it is underlined that 'nuclear safety and security are to a great extent scientific in nature, a regulatory body with responsibilities for nuclear facilities and other licensed activities involving radioactive material needs to have permanent access to supporting expertise'. This can be provided internally to the regulatory body or through an external entity such as a TSO in order to provide a competent and timely response to regulatory needs, and the development and maintenance of an appropriate knowledge base and associated tools and services, including training and tutoring services, are the main areas on which the conference will focus. This paper presents certain results obtained during carrying out the IAEA-Romanian Scientific Research Contract entitled: 'Romanian Contribution to the Development of Methodology for Risk Assessment and State Management of Nuclear Security Regime (Risk and Radiological Consequences)' where the author served as CSI (Chief Scientific Investigator) a part of the IAEA CRP-(Coordinated Research Project) on the: 'Development of Methodology for Risk Assessment and State Management of Nuclear Security Regime'. Also, potential risks and their possible radiological consequences to peoples and environment must not be underestimated. However, the assessment of risk must be realistic and quantified, and the requirements placed on the appropriate. The IAEA regulations for the Safe Transport of Radioactive Material, 2012 Edition, Specific Safety Requirements, SSR-6, are applicable to the national and international carriage of radioactive material by all modes of transport- by road, by rail, by air, by sea or by inland waters. The Radioactive Material to be transported must be categorized on the basis of its activity concentration, total activity, fissile characteristics (if any) and other relevant subsidiary characteristics. Packaging and package requirements are then specified on the basis of the hazard of the contents and range from normal commercial packaging (for low hazard contents) to strict design and performance requirements (for higher hazard contents). (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Control of Radioactive Methyl Iodide Gases by Mesoporous Hybrid Adsorbents at High Humid Condition - 19329

During the operation of nuclear facility as well as decommissioning stage, various radioactive gaseous wastes are generated. It is generally recognized that radioiodine is potentially one of the most hazardous gas on the human body and environment. Iodine-131 release in the gaseous effluents at the nuclear facilities had been a critical environmental concern. Reducing iodine release is a critical issue in radioactive waste system evaluation, especially in their adherence to the 'as low as reasonably achievable' (ALARA) criteria because radioiodine among these gaseous wastes has radiological harm. Therefore, numerous studies for the removal of radioiodine from the off-gas stream of nuclear facilities have been performed with formation processes and chemical structure. There has been some concern regarding the use of TEDA-impregnated activated carbon as an adsorbent for radioiodine removal, because of the low ignition point of TEDA and the possible release of TEDA from the charcoal in continuously flowing air streams. Therefore, the TEDA-impregnated activated carbon cannot use to trap radioiodine generated from high humid condition, due to the rapid drop of the removal efficiency. It is obvious that high humid can be expected during design basis event such as loss-of-coolant accident (LOCA). Therefore, investigation on the retention of organic radioiodine (CH3I131) by various activated carbons at high humid condition has extensively been performed so far. But there are no good results on the adsorption of methyl iodide at high humid condition up to now. As a result, the promising approach to the problem of iodine removal under high humid conditions is to use organic-inorganic hybrid adsorbents in which are enough retention characteristics from hydrophobic property. In order to develop the best removal efficiency for the control of radioiodine gas generated from a high humid process, a hydrophobic adsorbents was developed and the performance test of adsorbent using radioactive methyl iodide tracer was also performed. Demonstration test of adsorbent using radioactive methyl iodide tracer was performed at high humid conditions up to 97% relative humidity. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2023

RCT Module 2.12: Shipment and Receipt of Radioactive Material [Slides]

The basis of regulations governing the packaging and shipping of radioactive material is to keep radioactive material from affecting the environment during transportation and to keep the environment from affecting the integrity of the radioactive material. The package itself is to be designed and constructed to be the effective barrier between the environment and the radioactive material; thus, most of the regulatory restrictions apply to the package and the method of shipment used to transport the package. To reduce any potential hazard, the regulatory requirements become more restrictive as the quantity, concentration, and potential hazard of the radioactive material increase.

61 RADIATION PROTECTION AND DOSIMETRY↗

NASA NTRS2004

The Mean Life Squared Relationship for Abundances of Extinct Radioactivities

We discovered that the abundances of now extinct radioactivities (relative to stable reference isotopes) in meteorites vary as a function of their mean lifetimes squared. This relationship applies to chondrites, achondrites, and irons but to calcium-aluminum inclusions (CAIs). Certain meteorites contain excesses in isotopic abundances from the decay of radioactive isotopes with half-lives much less than the age of the solar system. These short-lived radioactivities are now extinct, but they were alive when meteorites assembled in the early solar system. The origin of these radioactivities and the processes which control their abundances in the solar nebula are still not well understood. Some clues may come from our finding that the meteoritic abundances of now extinct radioactivities (relative to stable reference isotopes) vary as a function of their mean lifetimes squared. This relationship applies to chondrites, achondrites, and irons, but not to CAIs. This points to at least two different processes establishing the abundances of short-lived isotopes found in the meteoritic record.

Lodders, K.↗

DOE OSTI2015

Methods of capturing and immobilizing radioactive nuclei with metal fluorite-based inorganic materials

Methods of capturing and immobilizing radioactive nuclei with metal fluorite-based inorganic materials are described. For example, a method of capturing and immobilizing radioactive nuclei includes flowing a gas stream through an exhaust apparatus. The exhaust apparatus includes a metal fluorite-based inorganic material. The gas stream includes a radioactive species. The radioactive species is removed from the gas stream by adsorbing the radioactive species to the metal fluorite-based inorganic material of the exhaust apparatus.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2018

Integrated Decision Methodology for Final Disposal of Radioactive Sources

Radioactive sources are used throughout the world, and their beneficial uses are many. Sources are at the core of many applications in different branches of industry, medicine, agriculture and research. Sources are present in a very wide range of equipment that is used for, inter alia, cancer treatment, killing bacteria in food, sterilizing medical supplies, measuring instruments such as gauges used to measure soil moisture and soil density and many other components, irradiating seeds for enhancing food production, protecting buildings from lightning strikes, mapping underground sources of water, prospecting for oil and gas reserves, measuring density of soil for construction projects, or even detecting smoke. At their “cradle” and when in use, radioactive sources are usually properly managed and controlled. It is when a source has reached the end of its useful life when they are at a higher risk of being under inadequate control, poorly managed, or even becoming orphan. Sources that have reached the end of their useful life must be carefully and safely managed. The “grave” of disused sources needs a proper disposal process to prevent them from posing a potential threat to people or the environment. To protect the public and the environment from the potential hazards of ionizing radiation, and to prevent disused sources from becoming orphan, a “cradle-to-grave” control of radioactive sources is essential, as promoted by the Code of Conduct on the Safety and Security of Radioactive Sources. This approach requires a national policy and strategy, an adequate legal and regulatory framework, and adequate resources and infrastructure. Many IAEAMember States have expressed the need for effective solutions to this challenge. There is also a gap related to the life cycle costs of potential disposal practices. This paper proposes a multi-attribute decision making methodology to compare disposal options using both economic and non-economic criteria leading to a preferred option for disposal and a strategy for final disposal of disused radioactive sources that is specific and optimized for each IAEA Member State.

Beatty, Randy↗

DOE OSTI2019

Radioactive Decay Computation with a Graph Approach for Solving a Linear Dynamic System

ABSTRACT: This article gives two graph-based algorithms for radioactive decay computation. The first algorithm identifies the connected components of the graph induced from the given radioactive decay dynamics to reduce the size of the problem. The solutions are derived over the precalculated connected components, respectively and independently. The second algorithm utilizes acyclic structure of radioactive decay dynamics. The algorithm evaluates the reachable vertices of the induced system graph from the initially activated vertices and finds the minimal set of starting vertices populating the entire reachable vertices. Then, the decay calculations are performed over the reachable vertices from the identified minimal starting vertices, respectively, with the partitioned initial value over the reachable vertices. A formal argument is given to show that the proposed graph inspired divide and conquer calculation methods perform the intended radioactive decay calculation. Empirical efforts comparing the proposed radioactive decay calculation algorithms are presented.

42 ENGINEERING↗

DOE OSTI2019

MULTI-MODAL CHALLENGES IN THE INTERNATIONAL TRANSPORT OF NUCLEAR AND RADIOACTIVE MATERIALS (THE ONLY INTERESTING PARTS OF A SYSTEM ARE THE INTERFACES.)

Transportation of nuclear and radioactive material exposes these materials to the most risk for malicious acts. Unlike physical facilities, securing radioactive and nuclear materials during transport requires that many different organizations including operators, regulatory agencies, and security agencies work together to ensure the security of the cargo. A major security challenge for international movement of nuclear and radioactive material is the need to use multiple transport modes. Ensuring security of these cargos as they move through intermodal facilities, some of which may take multiple hours, or days, depending on operational constraints at the yards, ports, or other facilities, is a major operational and regulatory challenge. Especially in the international context, ensuring the security of material in transport through in-transit countries and coordinating transport activities across jurisdictions can lead to challenges as to whose security plan controls, whose transport regulations are to be followed, and how might plans need to be adjusted to adapt to in-transit conditions and ultimately how to ensure cargo security. This paper introduces and highlights selected challenges in the multi-modal movement of nuclear and radioactive material in transport. While the paper will address selected challenges in coordination of stakeholders and resources, activities at intermodal facilities, and other selected issues, this does not address all the challenges, rather it brings issues forward and suggests solutions and strategies for enhancing the security of nuclear and radioactive material transported internationally.

Singley, Paul↗

DOE OSTI2019

Systems and methods for assaying a radioactive gas, and related scintillation cells

A radioactive gas assay system comprises a scintillation cell production assembly, a detector assembly, a computer assembly, and a scintillation cell destruction assembly. The scintillation cell production assembly is configured to produce a scintillation cell comprising a glass scintillator shell containing a volume of radioactive gas. The detector assembly is configured to receive the scintillation cell and to detect photons emitted thereby. The computer assembly is configured to receive data from the detector assembly to automatically calculate an absolute activity of the volume of radioactive gas of the scintillation cell and radiation detection efficiencies of the detector assembly. The scintillation cell destruction assembly is configured to receive the scintillation cell and to rupture the substantially non-porous glass scintillator shell to release the volume of radioactive gas. A method of assaying a radioactive gas, and a scintillation cell are also described.

Robinson, Troy A.↗