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Euratom performs safeguards monitoring of Fresh MOX fuel for domestic power production in the European Union. Video cameras monitor the reactor storage ponds. If video surveillance is lost for a certain amount of time a measurement is required to verify that no fuel was diverted. The attribute measurement to verify the continued presence of MOX fuel is neutron emission. Ideally this measurement would be made without moving or handling the fuel rod assembly. A prototype attribute measurement system was made using scintillating neutron sensitive glass waveguides developed by Pacific Northwest National Laboratory. Short lengths (5–20 cm) of the neutron sensitive fiber were mechanically spliced to 15 m lengths of commercial high numerical aperture fiber optic cable (Ceramoptec Optran Ultra 0.44). The light detector is a Hamamatsu R7400P photomultiplier tube. An electronics package was built to use the sensors with a GBS Elektronik MCA-166 multichannel analyzer and user interface. The MCA-166 is the system most commonly used by Euratom inspectors. It can also be run from a laptop computer using Maestro (Ortec) or other software. A MCNP model was made to compare to measurements made with several neutron sources including NIST traceable 252 Cf.

This paper presents an intuitive derivation of the source and absorber factors in a slab geometry formulation of γ-ray emissions incident on a detector. Using the attenuation equations, an application to mass estimation from a passively observed, low count spectrum is developed for sources with γ-ray emissions. Propagation of error provides a confidence interval on the mass estimate. The developed methods are applied to experimental spectra from a known mass of plutonium 239 oxide. The proposed estimation method has potential in national security applications where the geometry of attenuation materials is unknown and there are time constraints on the acquisition of a spectrum. In these application settings, the first-order mass estimate developed in this paper is sufficiently accurate.

A pulse shape analysis technique was implemented to determine electron mobility, /spl mu//sub e/, and electron mobility lifetime product, /spl mu//sub e//spl tau//sub e/ in cadmium zinc telluride detectors (CZT). The digital gamma finder, (DGF-4C), a single width CAMAC module produced by X-ray Instrumentation Associates (XIA), was used to extract pulse height, pulse shape, and signal rise time information. Post data acquisition analysis using the extracted information allowed measuring the /spl mu//sub e/ and /spl mu//sub e//spl tau//sub e/, in selected CZT samples. An almost linear relationship was observed for the signal rise time as a function of the inverse bias. This observed linear relationship was the basis for determination of /spl mu//sub e/ using a simple linear fit. The measured signal amplitude was also used to determine /spl mu//sub e//spl tau//sub e/ using the Hecht formulation. Repeated measurements carried out confirmed the consistency of the method in determining /spl mu//sub e/ and /spl mu//sub e//spl tau//sub e/.

High-resistivity crystals of CdZnTe are made using low-pressure Bridgman. The success of any doping scheme depends on its effectiveness in overcoming trace contamination and variations in processing. Multiple imaging and quantitative techniques gave highly complimentary information relating to spatial heterogeneity of the cadmium zinc telluride (CZT). Secondary Ion Mass Spectroscopy (SIMS) has good sensitivity and can quantitate trace elements in areas less than 150 μm in diameter. PNNL has SIMS standards for several key elements and can extrapolate for other elements. A heavily Pb-doped sample (0.1 atm%) was characterized to determine the distribution of lead and its potential impact on precipitates.

Since the mid-1980's the Pacific Northwest National Laboratory (PNNL) has used a value of 0.85 as a correction factor for the self absorption of activity for particulate radioactive air samples collected from building exhaust for environmental monitoring. More recently, an effort was made to evaluate the current particulate radioactive air sample filters (Versapor 3000, 47-mm diameter) used at PNNL for self absorption effects. There were two methods used to characterize the samples. Sixty samples were selected from the archive for acid digestion to compare the radioactivity measured by direct gas-flow proportional counting of filters to the results obtained after acid digestion of the filter and counting again by gas-flow proportional detection. Thirty different sample filters were selected for visible light microscopy to evaluate filter loading and particulate characteristics. Mass-loading effects were also considered. Large error is associated with the sample filter analysis comparison and subsequently with the estimation of the absorption factor resulting in an inadequate method to estimate losses from self-absorption in the sample filter. The mass loading on the sample filter as determined after digestion and drying was approximately 0.08 mg cm; however, this value may not represent the total filter mass loading given that there may be undetermined losses associated with the digestion process. While it is difficult to determine how much material is imbedded in the filter, observations from the microscopy analysis indicate that the vast majority of the particles remain on the top of the filter. In comparing the results obtained, the continued use of 0.85 as a conservative correction factor is recommended.

We present measured data on the neutron detection response of the GammaTracker handheld radioisotope identifier. Two neutron detection modes are discussed: measuring absorption gamma rays from the cadmium present in the Cd-Zn-Te spectrometers, and measuring the absorption gamma rays from moderator material present in the environment. In both cases, the capture gamma rays can be imaged to help locate a shielded neutron source. In this work, we discuss the total neutron detection efficiency of the GammaTracker instrument, and we present measured images of shielded neutron sources. Prospects for gamma-ray rejection are discussed.

A neutron detector designed for detecting neutron sources at distances of 50 to 100 m has been constructed and tested. This detector has a large surface area (1 m{sup 2}) to enhance detection efficiency, and it contains a collimator and shielding to achieve direction sensitivity and reduce background. An unusual feature of the detector is that it contains no added moderator, such as polyethylene, to moderate fast neutrons before they reach the {sup 3}He detector. As a result, the detector is sensitive mainly to thermal neutrons. The moderator-free design reduces the weight of the detector, making it more portable, and it also aids in achieving directional sensitivity and background reduction. Test results show that moderated fission-neutron sources of strength about 3 x 10{sup 5} n/s can be detected at a distance out to 70 m in a counting time of 1000 s. The best angular resolution of the detector is obtained at distances of 30 m or less. As the separation .distance between the source and detector increases, the contribution of scattered neutrons to the measured signal increases with a resultant decrease in the ability to detect the direction to a distant source. Applications for which the long-range detector appearsmore » to be suitable include detecting remote neutron sources (including sources in moving vehicles) and monitoring neutron storage vaults for the intrusion of humans and the effects they make on the detected neutron signal. Also, the detector can be used to measure waste for the presence of transuranic material in the presence of high gamma-ray background. A test with a neutron source (3 x 10{sup 5} n/s) in a vehicle showed that the detector could readily measure an increase in count rate at a distance of 10 m for vehicle speeds up to 35 mph (the highest speed tested). These results. indicate that the source should be detectable at this distance at speeds up to 55 mph.« less

The ability to collect broadband spectroscopic information about chemical analytes is highly desirable. We report on a technique that combines chemically selective coatings and optical spectroscopy. A 1-meter fiber 150 micrometers in diameter has approximately 5 cm2 surface area. This entire surface is used by incorporating selective moieties into the fiber cladding. The Large-Area Chemical Sensor concept for chemical sensing and measurement is based on a combination of three techniques. Specifically, it uses: (1) optical waveguides as the sensor substrate, (2) selectively adsorbing or absorbing materials to concentrate the target materials, and (3) spectroscopic interrogation for verification and quantification. The concept has been demonstrated for an iodine sensor by co-polymerizing methyl, phenyl siloxane into di-methyl siloxane. The phenyl group forms a charge-transfer complex with iodine which has an absorption at ca. 500 nm. Fused silica is the waveguide core. This system provides sensitivities in the 10-ppm range. The concept has been implemented into a prototype field iodine sensor unit. Work on the sensor concept continues with the goal of improving the sensitivity by allowing each photon multiple opportunities to interact with a target molecule.© (1999) COPYRIGHT SPIE--The International Society for Optical Engineering. Downloading of the abstract is permitted for personal use only.

Neutron-sensitive scintillating glass fiber sensors provide several advantages over 3 He and BF 3 gas-tubes for plutonium detection and surveillance. Large active areas provide significant improvements in sensitivity versus cost. In addition, the glass sensors offer a wide dynamic counting range, fast response time, and low microphonic susceptibility relative to conventional sensors. We report the results of detection limits for neutron glass panels used for portal, freight and vehicle monitoring.