Special Nuclear Material Research Articles

Comparison of LaBr3:Ce and NAI(Tl) scintillators for radio-isotope identification devices

Lanthanum bromide (LaBr3:Ce) scintillators offer significantly better resolution (<3 percent at 662keV) relative to sodium iodide (NaI(Tl)) but contain internal radioactivity that contributes to spectral counts. LaBr3:Ce has recently become available commercially in sizes large enough for the hand-held radio-isotope identification device (RIID) market. To study its potential for RIIDs, a series of measurements were performed comparing a 1.5×1.5-in. LaBr3:Ce detector with an Exploranium GR-135 RIID, which contains a 1.5×2.2-in. NaI(Tl) detector. Measurements were taken for short time frames and included examples of naturally occurring radioactive material (NORM), typically found in cargo, and special nuclear materials. To facilitate direct comparison, spectra from the different detectors were analyzed with the same isotope identification software (ORTEC ScintiVision™). In general, the LaBr3:Ce detector was able to find more peaks and find them faster than the NaI(Tl) detector. To the same level of significance, the LaBr3:Ce detector was usually two to three times faster. The notable exception was for 40K-containing NORM where interfering internal activity due to 138La in the LaBr3:Ce detector exists and NaI(Tl) consistently outperformed LaBr3:Ce.

Design and Testing of a Position-Sensitive Plastic Scintillator Detector for Fast Neutron Imaging

We describe the design and performance of a position sensitive scintillator detector developed for neutron measurements. Several of these detectors are to be used in the assembly of the Fast Neutron Imaging Telescope (FNIT), an instrument with imaging and energy measurement capabilities, sensitive to neutrons in the 2-20 MeV energy range. FNIT was initially conceived to study solar neutrons as a candidate instrument for the Inner Heliospheric Sentinels (IHS) program under formulation at NASA. It is now being adapted to locate Special Nuclear Material (SNM) for homeland security purposes by detecting fission neutrons and reconstructing the image of their source. The detection principle is based on multiple elastic neutron-proton scatterings in organic scintillator. The detector presented here utilizes wavelength-shifting (WLS) fibers, grooved into the plastic scintillator and read out by multianode photomultiplier tubes (MAPMTs) to determine scattering locations. By also measuring the recoil proton and scattered neutron's energies, the direction and energy spectrum of incident neutrons can be determined and discrete sources identified

Preliminary results utilizing high-energy fission product γ-rays to detect fissionable material in cargo

A concept for detecting the presence of special nuclear material (235U or 239Pu) concealed in intermodal cargo containers is described. It is based on interrogation with a pulsed beam of 7MeV neutrons that produce fission events and their β-delayed neutron emission or β-delayed high-energy γ radiation between beam pulses provide the detection signature. Fission product β-delayed γ-rays above 3MeV are nearly 10 times more abundant than β-delayed neutrons and are distinct from natural radioactivity and from nearly all of the induced activity in a normal cargo. Detector backgrounds and potential interferences with the fission signature radiation have been identified and quantified.

A kinematically beamed, low energy pulsed neutron source for active interrogation

We are developing a new active interrogation system based on a kinematically focused low energy neutron beam. The key idea is that one of the defining characteristics of special nuclear materials (SNM) is the ability for low energy or thermal neutrons to induce fission. Thus by using low energy neutrons for the interrogation source we can accomplish three goals: (1) energy discrimination allows us to measure the prompt fast fission neutrons produced while the interrogation beam is on; (2) neutrons with an energy of approximately 60–100keV do not fission 238U and Thorium, but penetrate bulk material nearly as far as high energy neutrons do and (3) below about 100keV neutrons lose their energy by kinematical collisions rather than via the nuclear (n,2n) or (n,n′) processes thus further simplifying the prompt neutron induced background. 60keV neutrons create a low radiation dose and readily thermal capture in normal materials, thus providing a clean spectroscopic signature of the intervening materials. The kinematically beamed source also eliminates the need for heavy backward and sideway neutron shielding. We have designed and built a very compact pulsed neutron source, based on an RFQ proton accelerator and a lithium target. We are developing fast neutron detectors that are nearly insensitive to the ever-present thermal neutron and neutron capture induced gamma ray background. The detection of only a few high energy fission neutrons in time correlation with the linac pulse will be a clear indication of the presence of SNM.

The "nuclear car wash": a scanner to detect illicit special nuclear material in cargo containers

There is an urgent need to improve the reliability of screening cargo containers for illicit nuclear material that may be hidden there for terrorist purposes. A screening system is described for the detection of fissionable material hidden in maritime cargo containers. The system makes use of a low-intensity neutron beam for producing fission and the detection of the abundant high-energy /spl gamma/ rays emitted in the /spl beta/-decay of short-lived fission products and /spl beta/-delayed neutrons. The abundance of the delayed /spl gamma/ rays is almost an order of magnitude larger than that of the delayed neutrons normally used to detect fission, and they are emitted on about the same time scale as the delayed neutrons, i.e., /spl sim/1 min. The energy and temporal distributions of the delayed /spl gamma/ rays provide a unique signature of fission. Because of their high energy, these delayed /spl gamma/ rays penetrate low-Z cargoes much more readily than the delayed neutrons. Coupled with their higher abundance, the signal from the delayed /spl gamma/ rays escaping from the container is predicted to be as much as six decades more intense than the delayed neutron signal, depending upon the type and thickness of the intervening cargo. The /spl gamma/ rays are detected in a large array of scintillators located along the sides of the container as it is moved through them. Measurements have confirmed the signal strength in somewhat idealized experiments and have also identified one interference when 14.5-MeV neutrons from the D, T reaction are used for the interrogation. The interference can be removed easily by the appropriate choice of the neutron source.

Comparison of PVT and NaI(Tl) scintillators for vehicle portal monitor applications

The demand for radiation portal monitor (RPM) systems has increased, and their capabilities are being further scrutinized as they are being applied to the task of detecting nuclear weapons, special nuclear material, and radiation dispersal device materials that could appear at borders. The requirements and constraints on RPM systems deployed at high-volume border crossings are significantly different from those at weapons facilities or steel recycling plants, where RPMs have been historically employed. In this new homeland security application, RPM systems must rapidly detect localized sources of radiation with a very high detection probability and low false-alarm rate, while screening all of the traffic without impeding the flow of commerce. In light of this new Department of Homeland Security application, the capabilities of two popular gamma-ray-detector materials as applied to these needs are re-examined. Both experimental data and computer simulations, together with practical deployment experience, are used to assess currently available polyvinyltoluene and NaI(Tl) gamma-ray detectors for border applications.

Comparison of CdTe and CdZnTe detectors for field determination of uranium isotopic enrichments

A performance comparison of a CdTe and a CdZnTe detector when exposed to uranium samples of various isotopic enrichments has been performed. These high-resolution detectors can assist in the rapid determination of uranium isotopic content of illicit material. Spectra were recorded from these room temperature semiconductor detectors with a portable multi-channel analyzer, both in the laboratory and in a field environment. Both detectors were operated below ambient temperature using the vendor supplied thermoelectric coolers. Both detectors had nominally the same active volume (18 mm3 for the CdZnTe and 25 mm3 for the CdTe detector) and resolution. Spectra of samples of known isotopic content were recorded at fixed geometries. An evaluation of potential signature g rays for the detection of enriched uranium was completed. Operational advantages and disadvantages of each detector are discussed. There is a need to improve the detection sensitivity during the interdiction of special nuclear materials (SNM) for increased homeland protection. It is essential to provide additional tools to first responders and law enforcement personnel for assessing nuclear and radiological threats.

Monte Carlo models for the production of β-delayed gamma-rays following fission of special nuclear materials

A Monte Carlo method for the estimation of β-delayed γ-ray spectra following fission is described that can accomodate an arbitrary time-dependent fission rate and photon collection history. The method invokes direct sampling of the independent fission yield distributions of the fissioning system, the branching ratios for decay of individual fission products and the spectral distributions for photon emission for each decay mode. Though computationally intensive, the method can provide a detailed estimate of the spectrum that would be recorded by an arbitrary spectrometer, and can prove useful in assessing the quality of evaluated data libraries, for identifying gaps in these libraries, etc. The method is illustrated by a first comparison of calculated and experimental spectra from decay of short-lived fission products following the reactions 235U(n th, f) and 239Pu(n th, f). For general purpose transport calculations, where detailed consideration of the large number of individual γ-ray transitions in a spectrum may be unnecessary, it is shown that an accurate and simple parameterization of a γ-ray source function can be obtained. These parametrizations should provide high-quality average spectral distributions that should prove useful in calculations describing photons escaping from thick attenuating media.

An application of gamma/X-ray/weighing methods for the complete in field assay of nuclear materials

The simultaneous determination of the isotopic vector and mass fraction from a single gamma/X-ray spectrum would represent an important step forward allowing the determination of the mass of both uranium and plutonium bearing materials more easily, reliably and accurately than existing NDA methods. The availability of a complete NDA assay on site would strongly reduce the necessity for shipment and further expensive analyses of samples, improve the timeliness of measurements and reduce the cost of Safeguards sample analyses. The paper describes a modular programming approach aiming at the implementation of a software package holding all different gamma/X-ray and weighing (GXW) subprograms for analysis of special nuclear materials (liquid U, solid U, liquid Pu, solid Pu, MOX).

Immobilized Low-Activity Waste Performance Assessment:2001 Version

ABSTRACTThe Hanford Immobilized Low-Activity Waste Performance Assessment[1] examines the longterm environmental and human health effects associated with the planned disposal of the vitrified low-activity fraction of waste presently contained in Hanford Site tanks. The tank waste is the byproduct of separating special nuclear materials from irradiated nuclear fuels over the past 50 years. This waste is stored in underground single- and double-shell tanks. The tank waste is to be retrieved, separated into low-activity and high-level fractions, and then immobilized by vitrification. The U.S. Department of Energy (DOE) plans to dispose of the low-activity fraction in the Hanford Site 200 East Area. The high-level fraction will be stored at the Hanford Site until a national repository is approved.This report provides the site-specific long-term environmental information needed by the DOE to modify the current Disposal Authorization Statement for the Hanford Site[2]. The original Disposition Authorization Statement was based on the 1998 version[3] of this performance assessment, which was conditionally accepted by DOE [4].The calculations in this performance assessment show that a “reasonable expectation” exists that the disposal of the immobilized low-level fraction of tank waste from the Hanford Site can meet environmental and health performance objectives. As shown by the sensitivity studies, this conclusion remains valid despite the conceptual designs of the disposal facility and the ILAW packaging having undergone changes.

Analysis of Nuclear Proliferation Resistance of DUPIC Fuel Cycle

This study compares the proliferation resistance of DUPIC (Direct Use of Spent PWR Fuel in CANDU) fuel cycle with other fuel cycle cases. The other fuel cycles considered in this study are PWR of once-through mode (PWR-OT), PWR of reprocessing mode (PWR-MOX), in which spent PWR fuel is reprocessed and recovered plutonium is used for making MOX (Mixed Oxide), CANDU with once-through mode (CANDU-OT), PWR fuel and CANDU fuel in a once- through mode with reactor grid equivalent to DUPIC fuel cycle (PWR-CANDU-OT). This study is focused on intrinsic barriers, especially, radiation field of the diverted material, which could be a significant accessibility barrier, amount of special nuclear material based on 1 GWe-yr that has to be diverted and the quality of the separated fissile material. It is indicated from plutonium analysis of each fuel cycle that the MOX spent fuel is containing the largest plutonium per MTHM but PWR-MOX option based on 1 GWe-yr has the best benefit in total plutonium consumption aspects. The DUPIC option is containing a little higher total plutonium based on 1 GWe-yr than the PWR-MOX case, but the DUPIC option has the lowest fissile plutonium content which could be another measure for proliferation resistance. On the whole, the CANDU-OT option has the largest fissile plutonium as well as total plutonium per GWe-yr, which means negative points in nuclear proliferation resistance aspects. It is indicated from the radiation field analysis that fresh DUPIC fuel could play an important radiation barrier role, more than even CANDU spent fuels. In conclusion, due to those inherent features, the DUPIC fuel cycle could include technical characteristics that comply naturally with the Spent Fuel Standard, at all steps along the DUPIC linkage between PWR and CANDU.

Detection sensitivity for special nuclear materials with an advanced high-pressure xenon detector and robust fitting analysis

The sensitivity for detection of two special nuclear materials, /sup 235/U and /sup 239/Pu, using an advanced high-pressure xenon detector (HPXe) and robust fitting analysis (RFA) with RobWin is experimentally determined for a set of collections with varying attenuation thicknesses of lead. These measurements provide essential information that enables comparisons with other detector systems for nuclear portal monitoring applications. Advantages of the combined use of HPXe and RFA for nuclear portal monitors are discussed, and recent progress with them is described briefly. An experiment designed to provide data from which models for nuclear portal monitor design can be grounded is described. Spectra were collected at CEA/Saclay with an HPXe unit from MEPhI for both /sup 235/U and /sup 239/Pu With attenuating thicknesses of lead ranging from zero to four millimeters. A single analysis procedure that yielded consistent results for all spectra was developed to simulate the automated application of analysis without user intervention and without advance knowledge of shielding thickness. These two spectrum series were then analyzed with RFA methodology for detection sensitivity and is reported in a form that can be used for comparison with other detector systems and for optimization of design parameters for HPXe/RFA portal monitor systems.

Development of high efficiency, multi-element CdZnTe detectors for portable measurement applications

We describe the development of detector arrays and electronics for large-volume, hand-held CdZnTe detectors with the same counting efficiency as portable NaI(Tl) detectors presently used for nuclear material measurement applications. The pulse-height resolution of the multi-element detectors is at least three times better than NaI(Tl) over a wide energy range (from 100 keV to several MeV), enabling more accurate measurements of gamma-rays emitted by special nuclear material. Arrays of up to eight coplanar grid detectors can be combined to make detectors ranging in size from 4 to 14 cm3. Because the number of spectroscopy channels is small, low-power, hand-held detectors can be manufactured with conventional printed circuit board technology, thus keeping the cost of multi-element detectors to a minimum. The design and performance of an 8-element detector is presented.

Portal, freight and vehicle monitor performance using scintillating glass fiber detectors for the detection of plutonium in the Illicit Trafficking Radiation Assessment Program

Neutron-sensitive scintillating glass fiber sensors provide several advantages over neutron-sensitive gas-tubes for plutonium detection and surveillance. Large area detectors and higher effective neutron capture density provide significant improvements in sensitivity versus cost. The glass sensors offer a wide dynamic counting range, fast response time, no transport hazard, greater operator safety and lower micro-phonic susceptibility relative to conventional 3He and 10BF3 sensors. We report on the testing methods used in Illicit Trafficking Radiation Assessment Program and our experience in evaluating the performance of the portal, freight and vehicle monitors for customs applications employing these new sensors. Results of false and real alarm tests and distribution tests are presented and compared to previous Monte Carlo and first principle calculations. These earlier calculations determined the detection limits for neutron glass panels used for continuous monitoring of special nuclear materials. Assumptions used in the model assumed a nominal background count rate for neutrons of 0.010 n·cm-2·s-1, an intrinsic neutron efficiency of 20% and an alarm set point of 5.2 sigma representing one false alarm per month. Measured neutron data taken with 252Cf and several gamma ray sources confirm the calculated data. Calculations assume an unshielded source and are conservative because they do not include environmental improvements in neutron economy. The results demonstrate that these systems are capable of reliably detecting special nuclearmaterials (SNM) and other radionuclides within a few seconds.

An explicit model-based diagnostic approach in a plutonium nitrate storage facility

Any movement of special nuclear material must be properly identified and quantified. An aid has been developed in support of this, which re-constructs the operational history of a storage facility from a limited amount of plant information. This re-construction is achieved via a quantitative computer simulation, where the operational history is represented as a set of events. The events are obtained by applying a combination of explicit-model-based reasoning with a simple change detector and with a knowledge-based approach. This approach enables alternative possibilities to be hypothesised and quantified. With the end-user in mind, emphasis is placed on the user interface.

Radiostrontium in the Savannah river site environment.

The Savannah River Site has produced plutonium, tritium, and other special nuclear materials for national defense, other government programs, and some civilian purposes. Strontium has been released to the environment during the operation of five reactors, two radiochemical processing facilities, and other supporting facilities. During the period 1954-1996, 110 GBq of 90Sr were released to the atmosphere and 5,400 GBq were released to site streams. The maximum individual effective dose equivalent at the site boundary was estimated to be 72 microSv from atmospheric releases and 25 microSv from liquid releases. The 80-km population dose was 0.3 person-Sv.

A Brief History of Nuclear Criticality Accidents in Russia—1953–1997

Fourteen nuclear criticality accidents that occurred in Russia between 1953 and 1997 are described. These accidents are significant because of the loss of control of special nuclear material and the resultant radiation doses to personnel, potential damage to equipment, and release of radioactive material to the workplace and the environment. A qualitative analysis of the causes and contributing factors to these accidents is presented along with a description of the radiation health effects to workers. The primary cause of most of these accidents was inadequate design that allowed the use of process equipment that did not preclude nuclear criticality on the basis of geometry. Personnel errors and violations of procedures were major contributing factors to these accidents.

Field applications of the Scout™ portable MCA

The use of Quantrad Sensor's ScoutTM in field type applications is described. The portability of the ScoutTM enables the user to obtain more accurate information in the field versus a survey meter. Isotopic identification is possible when ancillary information is combined with built-in software libraries. Data from the ScoutTM in remediation at Stanford Linear Accelerator (SLAC), NORM (Naturally Occurring Radioactive Maerial) measurements in California's Central Valley oil fields, medical isotope identification at nuclear pharmaceutical company and emergency response applications are presented. Additionally, custom software enabled the use of the ScoutTM in identification, qualification and detection of Special Nuclear Materials (SNM) in illicit trafficking and portal monitoring applications.

InSpector: A portable multi-purpose MCA

Decommissioning and decontamination work, safeguards related measurements for special nuclear materials, and many other in-situ applications require that the measuring system be taken to the sample and not the other way around. In addition to a portable detector, these types of measurements need a lightweight, rugged, battery operated MCA. Canberra's solution to this need is the versatile InSpector Multi-Channel Analyzer. This single instrument includes the MCA memory management, a high voltage power supply for either a NaI or a Ge detector, a spectroscopy grade amplifier, a digital stabilizer, and an ADC. It has a total weight of 3,2 kg, including batteries.

98/01702 Development of coprecipitated nickel-alumina catalyst for pyrocatalytic gasification of biomass: Garcia, L. et al. Biomass Energy Environ., Proc. Eur. Bioenergy Conf., 9th, 1996, 2, 1359–1364. Edited by Chartier, P., Elsevier, Oxford, UK

Associated particle imaging (API) is a powerful technique for special nuclear material (SNM) detection and characterization of fissile material configurations. A sealed-tube neutron generator has been under development by Lawrence Berkeley National Laboratory to reduce the beam spot size on the neutron production target to 1 mm in diameter for a several-fold increase in directional resolution and simultaneously increases the maximum attainable neutron flux. A permanent magnet 2.45 GHz microwave-driven ion source has been adopted in this time-tagged neutron source. This type of ion source provides a high plasma density that allows the use of a sub-millimeter aperture for the extraction of a sufficient ion beam current and lets us achieve a much reduced beam spot size on target without employing active focusing. The design of this API generator uses a custom-made radial high voltage insulator to minimize source to neutron production target distance and to provide for a simple ion source cooling arrangement. Preliminary experimental results showed that more than 100 μA of deuterium ions have been extracted, and the beam diameter on the neutron production target is around 1 mm.