Special Nuclear Material Research Articles

Detection of special nuclear material from delayed neutron emission induced by a dual-particle monoenergetic source

Detection of unique signatures of special nuclear materials is critical for their interdiction in a variety of nuclear security and nonproliferation scenarios. We report on the observation of delayed neutrons from fission of uranium induced in dual-particle active interrogation based on the 11B(d,n γ)12C nuclear reaction. Majority of the fissions are attributed to fast fission induced by the incident quasi-monoenergetic neutrons. A Li-doped glass–polymer composite scintillation neutron detector, which displays excellent neutron/γ discrimination at low energies, was used in the measurements, along with a recoil-based liquid scintillation detector. Time-dependent buildup and decay of delayed neutron emission from 238U were measured between the interrogating beam pulses and after the interrogating beam was turned off, respectively. Characteristic buildup and decay time profiles were compared to the common parametrization into six delayed neutron groups, finding a good agreement between the measurement and nuclear data. This method is promising for detecting fissile and fissionable materials in cargo scanning applications and can be readily integrated with transmission radiography using low-energy nuclear reaction sources.

Functional components for a design strategy: Hot cell shielding in the high reliability safeguards methodology

The high reliability safeguards (HRS) methodology has been established for the safeguardability of advanced nuclear energy systems (NESs). HRS is being developed in order to integrate safety, security, and safeguards concerns, while also optimizing these with operational goals for facilities that handle special nuclear material (SNM). Currently, a commercial pyroprocessing facility is used as an example system. One of the goals in the HRS methodology is to apply intrinsic features of the system to a design strategy. This current study investigates the thickness of the hot cell walls that could adequately shield processed materials. This is an important design consideration that carries implications regarding the formation of material balance areas, the location of key measurement points, and material flow in the facility.

The sciences and applications of the Electron LINAC-driven neutron source in Argentina

The Neutron Physics group at Centro Atomico Bariloche (CNEA, Argentina) has evolved for more than forty five years around a small 25MeV linear electron accelerator. It constitutes our compact accelerator-driven neutron source (CANS), which is dedicated to the use and development of neutronic methods to tackle problems of basic sciences and technological applications. Its historical first commitment has been the determination of the total cross sections of materials as a function of neutron energy by means of transmission experiments for thermal and sub-thermal neutrons. This also allowed testing theoretical models for the generation of scattering kernels and cross sections. Through the years, our interests moved from classic pulsed neutron diffraction, which included the development of high-precision methods for the determination of very low hydrogen content in metals, towards deep inelastic neutron scattering (DINS), a powerful tool for the determination of atomic momentum distribution in condensed matter. More recently non-intrusive techniques aimed at the scanning of large cargo containers have started to be developed with our CANS, testing the capacity and limitations to detect special nuclear material and dangerous substances. Also, the ever-present “bremsstrahlung” radiation has been recognized and tested as a useful complement to instrumental neutron activation, as it permits to detect other nuclear species through high-energy photon activation. The facility is also used for graduate and undergraduate students’ experimental work within the frame of Instituto Balseiro Physics and Nuclear Engineering courses of study, and also MSc and PhD theses work.

Initial results from a multiple monoenergetic gamma radiography system for nuclear security

The detection of assembled nuclear devices and concealed special nuclear materials (SNM) such as plutonium or uranium in commercial cargo traffic is a major challenge in mitigating the threat of nuclear terrorism. Currently available radiographic and active interrogation systems use ∼1–10MeV bremsstrahlung photon beams. Although simple to build and operate, bremsstrahlung-based systems deliver high radiation doses to the cargo and to potential stowaways. To eliminate problematic issues of high dose, we are developing a novel technique known as multiple monoenergetic gamma radiography (MMGR). MMGR uses ion-induced nuclear reactions to produce two monoenergetic gammas for dual-energy radiography. This allows us to image the areal density and effective atomic number (Zeff) of scanned cargo. We present initial results from the proof-of-concept experiment, which was conducted at the MIT Bates Research and Engineering Center. The purpose of the experiment was to assess the capabilities of MMGR to measure areal density and Zeff of container cargo mockups. The experiment used a 3.0MeV radiofrequency quadrupole accelerator to create sources of 4.44MeV and 15.11MeV gammas from the 11B(d,nγ)12C reaction in a thick natural boron target; the gammas are detected by an array of NaI(Tl) detectors after transmission through cargo mockups . The measured fluxes of transmitted 4.44MeV and 15.11MeV gammas were used to assess the areal density and Zeff. Initial results show that MMGR is capable of discriminating the presence of high-Z materials concealed in up to 30cm of iron shielding from low- and mid-Z materials present in the cargo mockup.

Characterization and simulation of soft gamma-ray mirrors for their use with spent fuel rods at reprocessing facilities.

The use of a grazing incidence optic to selectively reflect K-shell fluorescence emission and isotope-specific lines from special nuclear materials is a highly desirable nondestructive analysis method for use in reprocessing fuel environments. Preliminary measurements have been performed, and a simulation suite has been developed to give insight into the design of the x ray optics system as a function of the source emission, multilayer coating characteristics, and general experimental configurations. The experimental results are compared to the predictions from our simulation toolkit to illustrate the ray-tracing capability and explore the effect of modified optics in future measurement campaigns.

An inner-crystal neutron-scatter camera: Monte carlo simulation

Neutron energy emitted from special nuclear materials (SNMs) can be measured by using neutronproton scattering; also, the scattering angle can be calculated from the ratio of the scattered energy to the incident energy. By using position and energy information, we can image the original source position by using the backprojection and list-mode maximum-likelihood expectation maximization (MLEM) method. In this paper, we propose an inner-crystal neutron scatter camera system in which the detectors are not separated to obtain interactions at a variety of scatter angles; based on this system, we analyzed the characteristics of the corresponding neutron-scattering camera. The factors that affected the neutron-scatter image were the neutron velocity after scattering, the cut-off level of the time of flight (ToF), and the width of the cones used for image reconstruction. To determine the optimal point for the reconstruction of an image, we estimated the performance of the system by using the figure of merit (FoM). The optimal neutron-velocity (d/ToF) was ~0.3 × 107 m/s according to our simulation result, while the optimal cut-off level of the ToF was 4 ns as the latter minimized the noise while maintaining the required efficiency. The widths of the cones (e) also affected the full width at half maximum (FWHM) and the noise of the image. In terms of a simple source-geometry, whereby concepts such as “point source” were used, a large e value was suitable to achieve noise reduction; however, regarding the complicated source geometry, a small e value was favorable for precise reconstruction of the original source geometry for both the backprojection and the list-mode MLEM methods.

Commissioning and Field Tests of a Van-Mounted System for the Detection of Radioactive Sources and Special Nuclear Material

MODES_SNM project aimed at developing a mobile/portable modular detection system for radioactive sources and Special Nuclear Material (SNM). Its main goal was to deliver a tested prototype capable of passively detecting weak or shielded radioactive sources with accuracy higher than that of currently available systems. By the end of the project all the objectives have been successfully achieved. Results from the laboratory commissioning and the field tests are presented in this publication.

Uncovering Special Nuclear Materials by Low-energy Nuclear Reaction Imaging.

Weapons-grade uranium and plutonium could be used as nuclear explosives with extreme destructive potential. The problem of their detection, especially in standard cargo containers during transit, has been described as “searching for a needle in a haystack” because of the inherently low rate of spontaneous emission of characteristic penetrating radiation and the ease of its shielding. Currently, the only practical approach for uncovering well-shielded special nuclear materials is by use of active interrogation using an external radiation source. However, the similarity of these materials to shielding and the required radiation doses that may exceed regulatory limits prevent this method from being widely used in practice. We introduce a low-dose active detection technique, referred to as low-energy nuclear reaction imaging, which exploits the physics of interactions of multi-MeV monoenergetic photons and neutrons to simultaneously measure the material’s areal density and effective atomic number, while confirming the presence of fissionable materials by observing the beta-delayed neutron emission. For the first time, we demonstrate identification and imaging of uranium with this novel technique using a simple yet robust source, setting the stage for its wide adoption in security applications.

A high-energy, high-flux source of gamma-rays from all-optical non-linear Thomson scattering

γ-Ray sources are among the most fundamental experimental tools currently available to modern physics. As well as the obvious benefits to fundamental research, an ultra-bright source of γ-rays could form the foundation of scanning of shipping containers for special nuclear materials and provide the bases for new types of cancer therapy.However, for these applications to prove viable, γ-ray sources must become compact and relatively cheap to manufacture. In recent years, advances in laser technology have formed the cornerstone of optical sources of high energy electrons which already have been used to generate synchrotron radiation on a compact scale. Exploiting the scattering induced by a second laser, one can further enhance the energy and number of photons produced provided the problems of synchronisation and compact γ-ray detection are solved.Here, we report on the work that has been done in developing an all-optical and hence, compact non-linear Thomson scattering source, including the new methods of synchronisation and compact γ-ray detection. We present evidence of the generation of multi-MeV (maximum 16–18MeV) and ultra-high brilliance (exceeding 1020photonss−1mm−2mrad−2 0.1% BW at 15MeV) γ-ray beams. These characteristics are appealing for the paramount practical applications mentioned above.

Photofission product yields of 238U and 239Pu with 22-MeV bremsstrahlung

In homeland security and nuclear safeguards applications, non-destructive techniques to identify and quantify special nuclear materials are in great demand. Although nuclear materials naturally emit characteristic radiation (e.g. neutrons, γ-rays), their intensity and energy are normally low. Furthermore, such radiation could be intentionally shielded with ease or buried in high-level background. Active interrogation techniques based on photofission have been identified as effective assay approaches to address this issue. In designing such assay systems, nuclear data, like photofission product yields, plays a crucial role. Although fission yields for neutron-induced reactions have been well studied and readily available in various nuclear databases, data on photofission product yields is rather scarce. This poses a great challenge to the application of photofission techniques. In this work, short-lived high-energy delayed γ-rays from photofission of 238U were measured in between linac pulses. In addition, a list-mode system was developed to measure relatively long-lived delayed γ-rays from photofission of 238U and 239Pu after the irradiation. Time and energy information of each γ-ray event were simultaneously recorded by this system. Cumulative photofission product yields were then determined using the measured delayed γ-ray spectra.

Study of a 10B+ZnS(Ag) neutron detector as an alternative to 3He-based detectors in Homeland Security

The response of a scintillation neutron detector of ZnS(Ag) with 10B was calculated, using the MCNPX Monte Carlo Code. The detector consists of four panels of polymethyl methacrylate (PMMA) and five thin layers of ~0.017cm thick 10B+ZnS(Ag) in contact with the PMMA. The response was calculated for the bare detector and with different thicknesses of High Density Polyethylene, HDPE, moderator for 29 monoenergetic sources as well as 241AmBe and 252Cf neutrons sources. In these calculations the reaction rate 10B(n, α)7Li and the neutron fluence in the sensitive area of the detector 10B+ZnS(Ag) was estimated. Measurements were made at the Neutron Measurements Laboratory, Universidad Politécnica de Madrid, LMN-UPM, to quantify the detections in counts per second in response to a 252Cf neutron source separated 200cm. The MCNPX computations were compared with measurements to estimate the efficiency of ZnS(Ag) for detecting the α that is created in the 10B(n, α)7Li reaction. After validating new models with different geometries it will be possible to improve the detector response trying to achieve a sensitivity of 2.5cps-ng252Cf comparable with the response requirements for 3He detectors installed in the Radiation Portal Monitors, RPMs. This type of detector can be considered an alternative to the 3He detectors for detection of Special Nuclear Material, SNM.

Performance of an RPM Based on Gd-lined Plastic Scintillator for Neutron and Gamma Detection

INFN and ANN designed and developed a Radiation Portal Monitor (RPM) to detect gamma and neutron radiations using plastic scintillator coupled to gadolinium. The system is a portal for the screening of vehicle and cargo containers aiming at detecting the presence of radioactive elements and, in particular, special nuclear materials (SNM). The system was calibrated based on static measurements at the INFN laboratory and characterized with dynamic tests at the the JRC ITRAP+10 facility. Results showed that the system is not only compliant to international standards of its category, but exceeds them in most of the cases, demonstrating the validity of the novel technology and of the implemented layout.

A Comparative Study of Single-pulse and Double-pulse Laser-Induced Breakdown Spectroscopy with Uranium-containing Samples.

Laser-induced breakdown spectroscopy (LIBS) holds potential advantages in special nuclear material (SNM) sensing and nuclear forensics, which require rapid analysis, minimal sample preparation, and stand-off distance capability. SNM, such as U, however, result in crowded emission spectra with LIBS, and characteristic emission lines are challenging to discern. It is well-known that double-pulse LIBS (DPLIBS) improves the signal intensity for analytes over conventional single-pulse LIBS (SPLIBS). This study investigates the U signal in a glass matrix using DPLIBS and compares it to signal obtained using SPLIBS. Double-pulse LIBS involves sequential firing of a 1.06 µm Nd:YAG pre-pulse and 10.6 µm TEA CO2 heating pulse in a near collinear geometry. Optimization of experimental parameters including inter-pulse delay and energy follows identification of characteristic lines for the bulk analyte Ca and the minor constituent analyte U for both DPLIBS and SPLIBS. Spatial and temporal coupling of the two pulses in the proposed DPLIBS technique yields improvements in analytical merits with a negligible increase in damage to the sample compared to SPLIBS. Subsequently, the study discusses optimum plasma emission conditions of U lines and relative figures of merit in both SPLIBS and DPLIBS. Investigation into plasma characteristics also addresses plausible mechanisms related to the observed U analyte signal variation between SPLIBS and DPLIBS.

Spectroscopic neutron detection using composite scintillators

Shielded special nuclear material (SNM), especially highly enriched uranium, is exceptionally difficult to detect without the use of active interrogation (AI). We are investigating the potential use of low-dose active interrogation to realize simultaneous high-contrast imaging and photofission of SNM using energetic gamma-rays produced by low-energy nuclear reactions, such as [Formula: see text]B(d,n[Formula: see text]C and [Formula: see text]C(p,p[Formula: see text]C. Neutrons produced via fission are one reliable signature of the presence of SNM and are usually identified by their unique timing characteristics, such as the delayed neutron die-away. Fast neutron spectroscopy may provide additional useful discriminating characteristics for SNM detection. Spectroscopic measurements can be conducted by recoil-based or thermalization and capture-gated detectors; the latter may offer unique advantages since they facilitate low-statistics and event-by-event neutron energy measurements without spectrum unfolding. We describe the results of the development and characterization of a new type of capture-gated spectroscopic neutron detector based on a composite of scintillating polyvinyltoluene and lithium-doped scintillating glass in the form of millimeter-thick rods. The detector achieves >108 neutron–gamma discrimination resulting from its geometric properties and material selection. The design facilitates simultaneous pulse shape and pulse height discrimination, despite the fact that no materials intrinsically capable of pulse shape discrimination have been used to construct the detector. Accurate single-event measurements of neutron energy may be possible even when the energy is relatively low, such as with delayed fission neutrons. Simulation and preliminary measurements using the new composite detector are described, including those conducted using radioisotope sources and the low-dose active interrogation system based on low-energy nuclear reactions.

Testing the Delayed Gamma Capability in MCNP6

The mission of the Domestic Nuclear Detection Office is to quickly and reliably detect unauthorized attempts to import or transport special nuclear material for use against the United States. Developing detection equipment to meet this objective requires accurate simulation of both the detectable signature and detection mechanism.A delayed particle capability was initially added to MCNPX 2.6.A in 2005 to sample the radioactive fission product parents and emit decay particles resulting from the decay chain. To meet the objectives of detection scenario modeling, the capability was designed to sample a particular time for emitting particular multiplicity of a particular energy. Because the sampling process of selecting both time and energy is interdependent, to linearize the time and emission sampling, atom densities are computed at several discrete time steps, and the time-integrated production is computed by multiplying the atom density by the decay constant and time step size to produce a cumulative distribution function for sampling the emission time, energy, and multiplicity. The delayed particle capability was initially given a time-bin structure to help reasonably reproduce, from a qualitative sense, a fission benchmark by Beddingfield, which examined the delayed gamma emission. This original benchmark was only qualitative and did not contain the magnitudes of the actual measured data but did contain relative graphical representation of the spectra. A better benchmark with measured data was later provided by Hunt, Mozin, Reedy, Selpel, and Tobin at the Idaho Accelerator Center; however, because of the complexity of the benchmark setup, sizable systematic errors were expected in the modeling, and initial results compared to MCNPX 2.7.0 showed errors outside of statistical fluctuation.Presented here is a more simplified approach to benchmarking, utilizing closed form analytic solutions to the granddaughter equations for particular sets of decay systems. We examine five different decay chains (two-stage decay to stable) and show the predictability of the MCNP6 delayed gamma feature. Results do show that while the default delayed gamma calculations available in the MCNP6 1.0 release can give accurate results for some isotopes (e.g., 137Ba), the percent differences between the closed form analytic solutions and the MCNP6 calculations were often >40% (28Mg, 28Al, 42K, 47Ca, 47Sc, 60Co). With the MCNP6 1.1 Beta release, the tenth entry on the DBCN card allows improved calculation within <5% as compared to the closed form analytic solutions for immediate parent emissions and transient equilibrium systems. While the tenth entry on the DBCN card for MCNP6 1.1 gives much better results for transient equilibrium systems and parent emissions in general, it does little to improve daughter emissions of secular equilibrium systems. Hypotheses were presented as to why daughter emissions of secular equilibrium systems might be mispredicted in some cases and not in others.

NON-DESTRUCTIVE EXAMINATION USING NEUTRONS: A NUCLEAR WASTE AND ORPHANED SOURCE CHARACTERIZATION CASE STUDY APPLICABLE TO NUCLEAR FORENSICS

Found unknown radioactive material is often contained in a package so the chemical and physical form of the material itself is unknown, and the detail of the packaging is likewise unknown. Together, these present a significant risk on how to handle the package for destructive examination of its contents. Whether of nefarious origin or the result of less stringent practices of yesteryear the material needs to be properly identified and characterized for appropriate disposition. Results of neutron imaging, neutron diffraction, and delayed neutron analysis as applied to an examination of sealed capsules containing unknown radioactive materials are presented. The results demonstrate that neutron-based non-destructive examination techniques can be employed for inspecting encapsulated radioactive samples to identify the materials, to elucidate the internal physical structure of the radioactive material and encapsulation, and to estimate the mass of fissile and fissionable materials within. This characterization of orphaned radioactive special nuclear material illustrates the potential for these techniques in nuclear forensics investigations.

Combining Radiography and Passive Measurements for Radiological Threat Localization in Cargo

Detecting shielded special nuclear material (SNM) in a cargo container is a difficult problem, since shielding reduces the amount of radiation escaping the container. Radiography provides information that is complementary to that provided by passive gamma-ray detection systems: while not directly sensitive to radiological materials, radiography can reveal highly shielded regions that may mask a passive radiological signal. Combining these measurements has the potential to improve SNM detection, either through improved sensitivity or by providing a solution to the inverse problem to estimate source properties (strength and location). We present a data-fusion method that uses a radiograph to provide an estimate of the radiation-transport environment for gamma rays from potential sources. This approach makes quantitative use of radiographic images without relying on image interpretation, and results in a probabilistic description of likely source locations and strengths. We present results for this method for a modeled test case of a cargo container passing through a plastic-scintillator-based radiation portal monitor and a transmission-radiography system. We find that a radiograph-based inversion scheme allows for localization of a low-noise source placed randomly within the test container to within 40 cm, compared to 70 cm for triangulation alone, while strength estimation accuracy is improved by a factor of six. Improvements are seen in regions of both high and low shielding, but are most pronounced in highly shielded regions. The approach proposed here combines transmission and emission data in a manner that has not been explored in the cargo-screening literature, advancing the ability to accurately describe a hidden source based on currently-available instrumentation.

Image reconstruction of a neutron scatter camera

Recently, Sandia Laboratories developed a neutron scatter camera to detect special nuclear materials. This camera exhibits the following advantages: high efficiency, direction discrimination, neutron-gamma discrimination ability, and wide field of view. However, using the direct projection method, the angular resolution of this camera is limited by uncertainties in the energies estimated from pulse height and time of flight measurements. In this study, we established an eight-element neutron scatter camera and conducted the experiment with a 252Cf neutron source. The results show that it has an angular resolution better than 8° (1σ) and a detection efficiency of approximately 2.6×10-4. Using maximum likelihood expectation maximization method, the image artifact was eliminated, and the angular resolution was improved. We proposed an average scattering angle method to estimate the scattering energy of neutrons and Compton gamma rays. As such, we can obtain a recognizable image and energy spectrum of the source with some degradation of energy and image resolutions. Finally, a newly measured light response function based on the MPD-4 device was used for image reconstruction. Although we did not obtain a better result than that of the standard light response function, we have observed the effects of light response function on image reconstruction.

Demonstration of two-dimensional time-encoded imaging of fast neutrons

We present a neutron detector system based on time-encoded imaging, and demonstrate its applicability toward the spatial mapping of special nuclear material. We demonstrate that two-dimensional fast-neutron imaging with 2° resolution at 2m stand-off is feasible with only two instrumented detectors.

Verification of threshold activation detection (TAD) technique in prompt fission neutron detection using scintillators containing 19F

In the present study ⌀ 5′′× 3′′ and ⌀ 2′′× 2′′ EJ-313 liquid fluorocarbon as well as ⌀ 2′′ × 3′′ BaF2 scintillators were exposed to neutrons from a 252Cf neutron source and a Sodern Genie 16GT deuterium-tritium (D+T) neutron generator. The scintillators responses to β− particles with maximum endpoint energy of 10.4 MeV from the n+19F reactions were studied. Response of a ⌀ 5′′ × 3′′ BC-408 plastic scintillator was also studied as a reference. The β− particles are the products of interaction of fast neutrons with 19F which is a component of the EJ-313 and BaF2 scintillators. The method of fast neutron detection via fluorine activation is already known as Threshold Activation Detection (TAD) and was proposed for photofission prompt neutron detection from fissionable and Special Nuclear Materials (SNM) in the field of Homeland Security and Border Monitoring. Measurements of the number of counts between 6.0 and 10.5 MeV with a 252Cf source showed that the relative neutron detection efficiency ratio, defined as ϵBaF2 / ϵEJ−313−5′′, is 32.0% ± 2.3% and 44.6% ± 3.4% for front-on and side-on orientation of the BaF2, respectively. Moreover, the ⌀ 5′′ EJ-313 and side-on oriented BaF2 were also exposed to neutrons from the D+T neutron generator, and the relative efficiency ϵBaF2 / ϵEJ−313−5′′ was estimated to be 39.3%. Measurements of prompt photofission neutrons with the BaF2 detector by means of data acquisition after irradiation (out-of-beam) of nuclear material and between the beam pulses (beam-off) techniques were also conducted on the 9 MeV LINAC of the SAPHIR facility.