Nuclear Security Applications Research Articles

A high sensitivity 4π view gamma imager with a monolithic 3D position-sensitive detector

Abstract Imaging is an effective way to survey gamma radiation hotspots in nuclear safety, public health and homeland security applications. The existing gamma imagers either equip a collimator , which limits the imaging field of view (FOV) and sensitivity, or work without a collimator but require a high performance detector with sophisticated data processing and image reconstruction methods, e.g. the Compton camera. In this work, we demonstrate the feasibility of a high sensitivity 4 π view gamma imager design concept. Based on the photon event density distribution inside a monolithic 3D position-sensitive detector, the imager reconstructs the 4 π view of the surrounding gamma hotspots that the detector is exposed to. Monte Carlo simulations and experiments were conducted for several configurations of surrounding sparse gamma point sources in the energy range from 0.1 to 1.5 MeV. The results show that the reconstructed 4 π view image can identify multiple gamma hotspots, with the hotspots’ positioning accuracy and resolution dependent on the detector’s intrinsic resolution and the statistics of the acquired data. We conclude that the proposed design can be used for surveying gamma hotspots in nuclear security applications.

Fast neutron- and [formula omitted]-ray coincidence detection for nuclear security and safeguards applications

The use of passive and active interrogation techniques to evaluate materials concerning their content of special nuclear materials (SNM) is fundamental in fields such as nuclear safeguards and security. Detection of fast neutrons and γ rays, which are a characteristic signature of SNM, has several potential advantages compared with the commonly used systems based on thermal and epithermal neutron counters, the most important being the much shorter required coincidence times and the correspondingly reduced rate of background events due to accidental coincidences. Organic scintillators are well suited for this purpose due to their fast timing properties and composition being based on carbon and hydrogen with large elastic scattering cross-sections for fast neutrons. Organic scintillators also have suitable detection efficiency for γ rays and exhibit pulse shape properties which are favorable for distinguishing between neutrons and γ rays. This paper presents experimental results and Monte Carlo simulations for a neutron–neutron and γ-neutron coincidence detection setup for identification and characterization of SNM based on such detectors. The measurements were carried out on different samples of PuO2 material with varying content of 240Pu at the Joint Research Center (JRC) of the European Commission, Ispra, Italy. The results demonstrate significant advantages of fast neutron-γ coincidence detection over fast neutron–neutron coincidence counting for certain applications, e.g. for nuclear security systems, even in the presence of moderate amounts of shielding.

Synopsis of the Oak Ridge Radiation Protection Research Needs Workshop.

Radiation protection is foundational to harnessing the societal benefits of radiation in nuclear energy, security, and medicine applications. Significant challenges in radiation protection remain unaddressed for the nuclear fuel cycle, nuclear medicine, emergency response, national defense, and space exploration, as the United States is lacking a coherent research strategy prioritizing radiation protection mission needs and gaps in scientific knowledge to meet these needs. Research and development in the field of radiation protection calls for cooperation among governmental agencies, emergency responders, research organizations, and the academic community. Amidst atrophying national expertise in radiation protection, the Radiation Protection Research Needs Workshop was spearheaded by the Oak Ridge Associated Universities, Oak Ridge National Laboratory, and the Health Physics Society. This workshop facilitated critical dialogue among radiation stakeholders in the governmental and scientific communities, including national laboratories, academic institutions, and industry partners. The workshop featured presentations representing 12 federal agencies and breakout sessions involving the identification of scientific drivers by subject matter experts in each of the following areas: new fuel cycles/reactors, dosimetry, medical physics, instrumentation and operations, decontamination and decommissioning, space radiation, national defense, emergency response, environmental modeling, and low-dose effects. The goal of this workshop was to seek stakeholder input toward the development of a national strategic research agenda in the field of radiation protection. Consequently, the Health Physics Society has established a Special Task Force on Health Physics Research Needs, tasked with the prioritization of scientific drivers in radiation protection for the development of a national strategic research agenda.

A Numerical Investigation of Dielectric Laser Accelerators for Nuclear Security-Related Applications Using LSP

This paper investigates the potential use of dielectric laser accelerators (DLAs) for nuclear security applications. For the first time, the LSP code is used to look at this application. LSP is a fast running, parallelized particle-in-cell code that includes space charge effects. The focus of this paper is to ascertain the order of magnitude, amount of charge, that can be driven through the accelerator. From these results, a recommendation on the potential of DLAs for nuclear security applications is given.

Plug-and-play radiation sensor components for unmanned aerial system platform

Mobile radiation sensing techniques for surge deployment and wide area search are important for nuclear security applications. Radiation sensors were developed for the use with robotic platforms of unmanned aerial systems (UAS). Ambient temperature neutron and gamma CLYC sensor and high resolution CZT gamma sensor were integrated into the UAS as plug-and-play interchangeable components allowing easy attachment to a robot in field conditions. The time and position data were added via Robot Operating System to measured intensities of photon peaks and neutron rates enabling the input for the UAS flight control. The contour mapping of a radiation area by multiple UAS was simulated.

Evaluation of Eu:LiCAF for neutron detection utilizing SiPMs and portable electronics

With the increasing cost and decreasing availability of 3He, there have been many efforts to find alternative neutron detection materials. Lithium calcium aluminum fluoride (LiCAF) enriched to 95% 6Li doped with europium was evaluated here as a replacement material for 3He. Wafers 0.5 cm thick, consisting of LiCAF crystals in a rubberized matrix, were embedded with wavelength shifting fibers (WSF) and mated to silicon photo-multipliers (SiPMs) to measure the photon response in a flux of neutrons from a DD neutron generator. Excellent discrimination was realized between neutrons and gammas, and both pulse-height discrimination and pulse-shape analysis were explored. A Figure of Merit (FoM) of 1.03 was achieved. By applying pulse-shape analysis, a simple neutron count output was generated by utilizing a low-pass filter to suppress fast pulses from the SiPM output and subsequently applying a threshold to the remaining signal. Custom electronics were built to bias the SiPMs, then amplify, filter, discriminate, and digitize the LiCAF/WSF scintillation photons, resulting in a digital pulse that can easily be counted with any microcontroller or field programmable gate array. A significant advantage of LiCAF is that it can be fabricated into any shape/size (when embedded in a rubberized matrix), and the light output and transparency is sufficient to allow for thicker scintillators which enable detection of both thermal and epithermal neutrons. This work demonstrated that Eu:LiCAF is capable of discriminating gammas from neutrons and is a potential replacement material for 3He, especially for nuclear security applications and neutron spectroscopy.

Characterization of New-Generation Silicon Photomultipliers for Nuclear Security Applications

Silicon photomultipliers have received a great deal of interest recently for use in applications spanning a wide variety of fields, including nuclear safeguards and nonproliferation. For nuclear-related applications, the ability of silicon photomultipliers to discriminate neutrons from gamma rays using pulse shape discrimination when coupled with certain organic scintillators is a characteristic of utmost importance. This work reports on progress characterizing the performance of twenty different silicon photomultipliers from five manufacturers with an emphasis on pulse shape discrimination performance and timing. Results are presented on pulse shape discrimination performance as a function of overvoltage for 6-mm x 6-mm silicon photomultipliers, and the time response to stilbene is characterized for silicon photomultipliers of three different sizes. Finally, comparison with a photomultiplier tube shows that some new-generation silicon photomultipliers can perform as well as photomultiplier tubes in neutron-gamma ray discrimination.

Characterization of Large Volume CLYC Scintillators for Nuclear Security Applications

We report on our development of large volume Cs2LiYCl6 (CLYC) detectors for nuclear security applications. Three-inch diameter boules have been grown and 3-in right cylinders have been fabricated. Crystals containing either >95% 6Li or >99% 7Li have been grown for applications specific to thermal or fast neutron detection, respectively. We evaluated their gamma and neutron detection properties and the performance is as good as small size crystals. Gamma and neutron efficiencies were measured for large crystals and compared with smaller size crystals. With their excellent performance characteristics, and the ability to detect fast neutrons, CLYC detectors are excellent triple-mode scintillators for use in handheld and backpack instruments for nuclear security applications.

Neutron resonance analysis for nuclear safeguards and security applications

Neutron-induced reactions can be used to study the properties of nuclear materials of interest in the fields of nuclear safeguards and security. The elemental and isotopic composition of these materials can be determined by using the presence of resonance structures. This idea is the basis of two non-destructive analysis techniques which have been developed at the GELINA neutron time-of-flight facility at JRC-Geel: Neutron Resonance Capture Analysis (NRCA) and Neutron Resonance Transmission Analysis (NRTA). A combination of NRTA and NRCA has been proposed for the characterisation of particle-like debris of melted fuel formed in severe nuclear accidents. In this work, we present a quantitative validation of the NRTA technique which was used to determine the areal densities of Pu enriched reference samples used for safeguards applications. Less than 2% bias has been obtained for the fissile isotopes, with well-known total cross sections.

Determination of relative krypton fission product yields from 14 MeV neutron induced fission of 238U at the National Ignition Facility.

Precisely-known fission yield distributions are needed to determine a fissioning isotope and the incident neutron energy in nuclear security applications. 14 MeV neutrons from DT fusion at the National Ignition Facility induce fission in depleted uranium contained in the target assembly hohlraum. The fission yields of Kr isotopes (85m, 87, 88, and 89) are measured relative to the cumulative yield of 88Kr and compared to previously tabulated values. The results from this experiment and England and Rider are in agreement, except for the 85mKr/88Kr ratio, which may be the result of incorrect nuclear data.

Intrinsic radioactivity of KSr2I5:Eu2+

A current need in nuclear security is an economical, yet high energy resolution (near 2%), scintillation detector suitable for gamma-ray spectroscopy. For current scintillators on the market, there is an inverse relationship between scintillator energy resolution and cost of production. A new promising scintillator, KSr2I5:Eu2+, under development at the University of Tennessee, has achieved an energy resolution of 2.4% at 662keV at room temperature, with potential growth rates exceeding several millimeters per hour. However, the internal background due to the 40K content could present a hurdle for effective source detection/identification in nuclear security applications. As a first step in addressing this question, this paper reports on a computational investigation of the intrinsic differential pulse height spectrum (DPHS) generated by 40K within the KSr2I5:Eu2+ scintillator as a function of crystal geometry. It was found that the DPHS remains relatively equal to a constant multiplicative factor of the negatron emission spectrum with a direct increase of the 1.46MeV photopeak relative height to the negatron spectrum with volume. Further, peak pileup does not readily manifest itself for practical KSr2I5:Eu2+ volumes.

Examination of spent fuel radiation energy conversion for electricity generation

Supply of electricity inside nuclear power plant is one of the most important considerations for nuclear safety and security. In this study, generation of electric energy by converting radiation energy of spent nuclear fuel was investigated. Computational modeling work by using MCNPX 2.7.0 code along with experiment was performed to estimate the amount of electric energy generation. The calculation using the developed modeling work was validated through comparison with an integrated experiment. The amount of electric energy generation based on a conceptual design of an energy conversion module was estimated to be low. But the amount may be useful for nuclear security applications. An alternative way of utilizing the produced electric energy could be considered for nuclear safety application through energy storage. Further studies are needed to improve the efficiency of the proposed energy conversion concept and to examine the issue of radiation damage and economic feasibility.

Enabling in situ thermometry using transmission nuclear resonance fluorescence

Transmission nuclear resonance fluorescence (NRF) has been proposed for use as an assay and detection technique for nuclear security and safeguards applications because of its isotope-specific sensitivity and the penetrating capability of γ-rays. It can also be used for in situ thermometry because the absorbing resonance profile is sensitive to temperature. Using transmission NRF for thermometry could provide a new avenue for studying the ion temperature evolution of laser-induced plasmas using the upcoming ELI-NP facility. It could also be used for applications where thermometry would be otherwise infeasible, such as for determining the average fuel temperature of spent nuclear fuel, a step that would reduce the assay uncertainty using transmission NRF. In this paper, two different transmission NRF thermometry scenarios are presented and analyzed for sensitivity. This analysis demonstrated that thermometry using γ-rays is possible independent of γ-ray beam type and will be feasible with next-generation high-intensity γ-ray sources. Beyond thermometry, an application of temperature effects was found in the improvement of transmission NRF efficacy for assay and detection: using a cryogenic witness target will reduce the required measurement time by 40%.

Magnetic imaging: a new tool for UK national nuclear security.

Combating illicit trafficking of Special Nuclear Material may require the ability to image through electromagnetic shields. This is the case when the trafficking involves cargo containers. Thus, suitable detection techniques are required to penetrate a ferromagnetic enclosure. The present study considers techniques that employ an electromagnetic based principle of detection. It is generally assumed that a ferromagnetic metallic enclosure will effectively act as a Faraday cage to electromagnetic radiation and therefore screen any form of interrogating electromagnetic radiation from penetrating, thus denying the detection of any eventual hidden material. In contrast, we demonstrate that it is actually possible to capture magnetic images of a conductive object through a set of metallic ferromagnetic enclosures. This validates electromagnetic interrogation techniques as a potential detection tool for National Nuclear Security applications.

Correlated statistical uncertainties in coded-aperture imaging

In nuclear security applications, coded-aperture imagers can provide a wealth of information regarding the attributes of both the radioactive and nonradioactive components of the objects being imaged. However, for optimum benefit to the community, spatial attributes need to be determined in a quantitative and statistically meaningful manner. To address a deficiency of quantifiable errors in coded-aperture imaging, we present uncertainty matrices containing covariance terms between image pixels for MURA mask patterns. We calculated these correlated uncertainties as functions of variation in mask rank, mask pattern over-sampling, and whether or not anti-mask data are included. Utilizing simulated point source data, we found that correlations arose when two or more image pixels were summed. Furthermore, we found that the presence of correlations was heightened by the process of over-sampling, while correlations were suppressed by the inclusion of anti-mask data and with increased mask rank. As an application of this result, we explored how statistics-based alarming is impacted in a radiological search scenario.

Applicability of Raman spectroscopy as a tool in nuclear forensics for analysis of uranium ore concentrates

Abstract In the present study, the suitability of Raman spectroscopy for nuclear forensics is proven by a systematic investigation of laboratory-prepared synthetic and industrial uranium ore concentrate (UOC) samples, of different origin and with wide range of impurity content. Besides the identification of the compound type, the identification of process-related impurities observed in these samples is reported for the first time. Both handheld and laboratory Raman spectrometers were used in the measurements. The ability to measure different UOCs and the sufficient sensitivity of the handheld device demonstrate its usefulness in nuclear safeguards and nuclear security applications. Even more importantly in such a context, the likely identity of process-related impurities could be verified by Raman Spectroscopy in the present work, allowing to draw conclusions on the process history of the sample.

Pareto-Optimal Gamma Spectroscopic Radionuclide Identification Using Evolutionary Computing

Detection and identification of special nuclear materials (SNM) in measured gamma spectra is a challenge for nuclear security applications of spectroscopy. There are ongoing international efforts in industry, academia, and government to develop accurate, robust automated methods of gamma spectroscopic analysis to identify signature patterns characteristic of SNM. In this paper, we introduce a new approach for processing gamma-ray spectra measured by an NaI detector. The approach searches for radionuclide signature patterns in the measured spectrum by formulating a multiobjective optimization problem. Search in the multiobjective space is performed by an evolutionary algorithm, which identifies a solution in the context of Pareto Optimality Theory. The proposed approach is composed of two main steps: 1) signature selection, and 2) Pareto-optimal fitting. The output of the algorithm includes a list of all the identified radionuclides and their relative contributions to the gamma spectrum. The methodology is applied to a set of gamma spectra and compared to multiple linear regression fitting. Results demonstrate the superiority of the Pareto-optimal approach over multiple regression in the majority of the tested cases.

A proposed benchmark problem for cargo nuclear threat monitoring

Abstract There is currently a great deal of technical and political effort focused on reducing the risk of potential attacks on the United States involving radiological dispersal devices or nuclear weapons. This paper proposes a benchmark problem for gamma-ray and X-ray cargo monitoring with results calculated using MCNP5, v1.51. The primary goal is to provide a benchmark problem that will allow researchers in this area to evaluate Monte Carlo models for both speed and accuracy in both forward and inverse calculational codes and approaches for nuclear security applications. A previous benchmark problem was developed by one of the authors (RPG) for two similar oil well logging problems (Gardner and Verghese, 1991, [1] ). One of those benchmarks has recently been used by at least two researchers in the nuclear threat area to evaluate the speed and accuracy of Monte Carlo codes combined with variance reduction techniques. This apparent need has prompted us to design this benchmark problem specifically for the nuclear threat researcher. This benchmark consists of conceptual design and preliminary calculational results using gamma‐ray interactions on a system containing three thicknesses of three different shielding materials. A point source is placed inside the three materials lead, aluminum, and plywood. The first two materials are in right circular cylindrical form while the third is a cube. The entire system rests on a sufficiently thick lead base so as to reduce undesired scattering events. The configuration was arranged in such a manner that as gamma-ray moves from the source outward it first passes through the lead circular cylinder, then the aluminum circular cylinder, and finally the wooden cube before reaching the detector. A 2 in.×4 in.×16 in. box style NaI (Tl) detector was placed 1 m from the point source located in the center with the 4 in.×16 in. side facing the system. The two sources used in the benchmark are 137 Cs and 235 U.

Evaluation of large deuterated scintillators for fast neutron detection (E=0.5–20MeV) using the D(d,n)3He, 13C(d,n) and 27Al(d,n) reactions

Abstract The response of large deuterated liquid scintillators (up to 10 cm diameter by 15 cm) to neutrons 0.5–20 MeV has been studied using the 2.5 MeV neutron generator at the University of Michigan, and the d(d,n), 13C(d,n), 27Al(d,n) and other reactions at the University of Notre Dame FN tandem accelerator. The latter utilize 9 and 16 MeV deuteron beams including a pulsed beam, which permitted time-of-flight measurements. Combining pulse-shape discrimination and time-of-flight allows gating on specific neutron energy groups to determine the detector response to specific neutron energies. This will permit accurate simulation of the detector response functions for applications of these detectors in nuclear research and homeland security applications.

Detection of SNM by delayed gamma rays from induced fission

The Pulsed Neutron Interrogation Test Assembly (PUNITA) is an experimental device for research in NDA methods and field applicable instrumentation for nuclear safeguards and security applications. PUNITA incorporates a standard 14-MeV (D-T) pulsed neutron generator inside a large graphite mantle. The generator target is surrounded by a thick tungsten filter with the purpose to increase the neutron output and to tailor the neutron energy spectrum . In this configuration a sample may be exposed to a relatively high average thermal neutron flux of about (2.2±0.1)×10 3 s −1 cm −2 at only 10% of the maximum target neutron emission . The sample cavity is large enough to allow variation of the experimental setup including the fissile sample, neutron and gamma detectors , and shielding materials. The response from SNM samples of different fissile material content was investigated with various field-applicable scintillation gamma detectors such as the 3×2 in. LaBr 3 detector. Shielding in the form of tungsten and cadmium was applied to the detector to improve the signal to background ratio. Gamma and neutron shields surrounding the samples were also tested for the purpose of simulating clandestine conduct. The energy spectra of delayed gamma rays were recorded in the range 100 keV–9 MeV. In addition time spectra of delayed gamma rays in the range 3.3–8 MeV were recorded in the time period of 10 ms–120 s after the 14-MeV neutron burst. The goal of the experiment was to optimize the sample/detector configuration including the energy range and time period for SNM detection. The results show, for example, that a 170 g sample of depleted uranium can be detected with the given setup in less than 3 min of investigation. Samples of higher enrichment or higher mass are detected in much shorter time.