John E. Smart

Field tests of a NaI(Tl)-based vehicle portal monitor at border crossings

Radiation portal monitors are commonly used at international border crossings to detect illicit transport of radioactive material. Most monitors use plastic scintillators to detect gamma rays, but next-generation monitors may contain NaI(Tl). In order to directly compare the performance of the two types of detectors, a prototype NaI(Tl) monitor was tested at two international border crossings adjacent to a comparable plastic scintillator monitor. The NaI(Tl) monitor housed four large detectors, each 10.2 cm /spl times/ 10.2 cm /spl times/ 41 cm. The empirical data set from the two field tests contains approximately 3800 passages with known cargo loads for each vehicle. For a small subset of the vehicles, high purity germanium detector spectra were also collected. During the survey period several vehicles containing commercial products with naturally occurring radioactive material (NORM) passed through the monitor. Typical NORM cargo included pottery, large granite slabs, rock-based floor tiles, construction stone blocks, abrasive material, and fertilizer. Non-NORM sources included a large source of /sup 60/Co (200,000 GBq) and a shipment of uranium oxide, both items being legally transported. The information obtained during the tests provides a good empirical data set to compare the effectiveness of NaI(Tl) and plastic-scintillator portal monitors. The capability to be sensitive to illicit materials, but not alarm on NORM, is a key figure of merit for portal monitors.

System Modeling and Design Optimization for a Next-Generation Unattended Sensor

We are developing a next-generation unattended sensor that can detect and identify radiation sources while operating on battery power for several weeks. The system achieves smaller size and weight over systems that use NaI:Tl and 3He detectors by using a relatively new scintillator, Cs2LiYCl6:Ce (CLYC). This material can detect both gamma rays and thermal neutrons, has best-case energy resolution under 4% full width at half maximum at 662 keV, and allows for particle discrimination by pulse amplitude as well as pulse shape. The overall design features an array of sixteen CLYC detectors, each read out by a photomultiplier tube and custom pulse processing electronics. A field-programmable gate array analyzes the energy spectra using computationally efficient algorithms for anomaly detection and basic isotope identification. In this paper, we report the results of a modeling study to optimize various parameters of the unattended sensor for best performance. Key parameters include the number and placement of detectors, dimensions and weight of the moderator, and location of the batteries. These results have guided the design of the proof-of-concept prototype.

Using Atmospheric Dispersion Theory to Inform the Design of a Short-lived Radioactive Particle Release Experiment.

AbstractAtmospheric dispersion theory can be used to predict ground deposition of particulates downwind of a radionuclide release. This paper uses standard formulations found in Gaussian plume models to inform the design of an experimental release of short-lived radioactive particles into the atmosphere. Specifically, a source depletion algorithm is used to determine the optimum particle size and release height that maximizes the near-field deposition while minimizing both the required source activity and the fraction of activity lost to long-distance transport. The purpose of the release is to provide a realistic deposition pattern that might be observed downwind of a small-scale vent from an underground nuclear explosion. The deposition field will be used, in part, to study several techniques of gamma radiation survey and spectrometry that could be used by an On-Site Inspection team investigating such an event.

PRex: An Experiment to Investigate Detection of Near-field Particulate Deposition from a Simulated Underground Nuclear Weapons Test Vent.

AbstractA radioactive particulate release experiment to produce a near-field ground deposition representative of small-scale venting from an underground nuclear test was conducted to gather data in support of treaty capability development activities. For this experiment, a CO2‐driven “air cannon” was used to inject 140La, a radioisotope of lanthanum with 1.7‐d half-life and strong gamma-ray emissions, into the lowest levels of the atmosphere at ambient temperatures. Witness plates and air samplers were laid out in an irregular grid covering the area where the plume was anticipated to deposit based on climatological wind records. This experiment was performed at the Nevada National Security Site, where existing infrastructure, radiological procedures, and support personnel facilitated planning and execution of the work. A vehicle-mounted NaI(Tl) spectrometer and a polyvinyl toluene-based backpack instrument were used to survey the deposited plume. Hand-held instruments, including NaI(Tl) and lanthanum bromide scintillators and high purity germanium spectrometers, were used to take in situ measurements. Additionally, three soil sampling techniques were investigated and compared. The relative sensitivity and utility of sampling and survey methods are discussed in the context of on-site inspection.

Unattended Sensor System With CLYC Detectors

We have developed an unattended sensor for detecting anomalous radiation sources. The system combines several technologies to reduce size and weight, increase battery lifetime, and improve decision-making capabilities. Sixteen Cs2LiYCl6:Ce (CLYC) scintillators allow for gamma-ray spectroscopy and neutron detection in the same volume. Low-power electronics for readout, high voltage bias, and digital processing reduce the total operating power to 1.7 W. Computationally efficient analysis algorithms perform spectral anomaly detection and isotope identification. When an alarm occurs, the system transmits alarm information over a cellular modem. In this paper, we describe the overall design of the unattended sensor, present characterization results, and compare the performance to stock NaI:Tl and 3He detectors.

Particle Release Experiment (PRex) Final Report

...................................................................................................................................................... iii Executive Summary .................................................................................................................................... v Acknowledgments ..................................................................................................................................... vii Acronyms and Abbreviations ................................................................................................................... ix 1.0 Introduction ............................................................................................................................ 1 1.1 Background .............................................................................................................................. 1 1.2 Objective .................................................................................................................................. 2 1.3 Scope of Document .................................................................................................................. 2 2.0 Experimental Considerations ................................................................................................ 3 2.1 Overview of PRex .................................................................................................................... 3 2.2 Constraints ............................................................................................................................... 3 2.3 Assumptions ............................................................................................................................. 4 2.4 Pre-Execution Release Parameters ........................................................................................... 4 3.0 PRex Experiment Preparation .............................................................................................. 6 3.1 Experiment Plan ....................................................................................................................... 6 3.2 Site Selection............................................................................................................................ 6 3.3 Background Measurements, September and October 2012 ..................................................... 7 3.4 Source Production .................................................................................................................... 9 3.4.1 Preparation of La2O3 Powder .................................................................................................................. 9 3.4.2 Irradiation at Washington State University ........................................................................................... 12 3.4.3 Source Transport ................................................................................................................................... 14 3.5 Release Mechanism ................................................................................................................ 19 3.6 Meteorology Planning and Modeling ..................................................................................... 22 3.7 NNSS Experiment Site Preparation ....................................................................................... 22 3.7.1 Posting Contamination and High Contamination Areas ....................................................................... 22 3.7.2 Prepositioned Sampling ........................................................................................................................ 22 4.0 PRex Experiment Execution ............................................................................................... 28 4.1 Time and Location ................................................................................................................. 28 4.2 Meteorological Conditions during Release ............................................................................ 28 4.3 Sampling Network ................................................................................................................. 32 4.3.1 Collection of Witness Plates ................................................................................................................. 32 4.3.2 Soil Sampling ....................................................................................................................................... 33 4.3.3 Air Sampling ......................................................................................................................................... 36 4.4 Radiation Survey .................................................................................................................... 36 4.4.1 Vehicle Survey...................................................................................................................................... 37 4.4.2 Handheld/Man-portable ........................................................................................................................ 38 4.4.3 Aerial Survey ........................................................................................................................................ 39 4.5 Field Laboratory ..................................................................................................................... 40

A neutron monitor for in-situ real-time determination of transuranics in a processed waste effluent stream

A pilot plant operation at the Savannah River Site will remove 90Sr, 137Cs, and transuranics from a high-level liquid waste stream prior to encapsulation in a Saltstone Facility. Monitors are required to determine the concentrations of all radionuclides, including transuranics, in real-time on this processed waste stream. A neutron counter used to measure the concentration of each actinide isotope present in the stream is described. The neutron counter assembly consists of nested annular layers of shielding, reflectors, detectors, and moderators. On-line, live-time system control and calibration is provided by a time-tagged neutron source embedded in the moderator assembly.

Calibration of an automated high resolution gamma-ray spectroscopy system for in-situ real-time monitoring of a processed waste effluent stream

A pilot plant is being designed at the U. S. Department of Energys Savannah River Site (SRS) to demonstrate the removal of 90Sr, 137Cs, and transuranics from a high-level liquid waste stream prior to encapsulation in a Saltstone Facility. In-line monitors are required to determine the concentration of all radionuclides on this processed waste stream. Calibration standards containing 60Co, 137Cs, and 90Sr were prepared and counted. Efficiency curves were generated. Strontium-90 is readily observable above the system background in the calibration standard count, and is observable at less than 3 nCi/ml in a mixed solution having the maximum allowable concentration of all other activities present in the proposed SRS effluent stream.