Nondestructive Assay Techniques Research Articles

Compact and transportable system for detecting lead-shielded highly enriched uranium using 252Cf rotation method with a water Cherenkov neutron detector

The global challenge of on-site detection of highly enriched uranium (HEU), a substance with considerable potential for unauthorized use in nuclear security, is a critical concern. Traditional passive nondestructive assay (NDA) techniques, such as gamma-ray spectroscopy with high-purity germanium detectors, face significant challenges in detecting HEU when it is shielded by heavy metals. Addressing this critical security need, we introduce an on-site detection method for lead-shielded HEU employing a transportable NDA system that utilizes the 252Cf rotation method with a water Cherenkov neutron detector. This cost-effective NDA system is capable of detecting 4.17 g of 235U within a 12 min measurement period using a 252Cf source of 3.7 MBq. Integrating this system into border control measures can enhance the prevention of HEU proliferation significantly and offer robust deterrence against nuclear terrorism.

Numerical intercomparison of PHITS and Geant4 Monte Carlo codes for fast neutron inelastic scattering applications

Fast neutron inelastic scattering is a promising non-destructive assay technique for various analytical applications. As an active neutron interrogation technique, its performance is a function of various different factors and parameters that require optimization. Monte Carlo simulation codes are indispensable for such tasks. However, the internal simulation routines implemented in such codes can rely on different physical models that can yield discrepancies in the simulation results. In this work we conduct an intercomparison of PHITS and Geant4 codes performance in application to fast neutron inelastic scattering simulations. The goal of this paper is twofold. First, we explain the differences in code configuration with respect to gamma and neutron transport, as well as internal simulation routines. Second, we conduct a performance assessment of the two codes using two different measurement configurations. One configuration consisted of a source of gamma rays in a broad energy range (100–9000 keV) and a CeBr3 detector. The other configuration consisted of a monoenergetic 2.5 MeV fast neutron source, Fe, Nd, Dy, B targets and a CeBr3 detector. Selected simulation configurations were chosen with a goal to compare the performance differences in neutron energy distribution, produced prompt gamma rays and energy deposition in CeBr3 detector between the two codes. Results of our study reveal a good coherence of both codes performance in the application of fast neutron inelastic scattering simulations. The simulation geometries and observed differences are described in detail.

Multiplicity counting using organic scintillators to distinguish neutron sources: An advanced teaching laboratory

In this advanced instructional laboratory, students explore complex detection systems and nondestructive assay techniques used in the field of nuclear physics. After setting up and calibrating a neutron detection system, students carry out timing and energy deposition analyses of radiation signals. Through the timing of prompt fission neutron signals, multiplicity counting is used to carry out a special nuclear material (SNM) nondestructive assay. Our experimental setup is comprised of eight trans-stilbene organic scintillation detectors in a well-counter configuration, and measurements are taken on a spontaneous fission source as well as two (α,n) sources. By comparing each source's measured multiplicity distribution, the resulting measurements of the (α,n) sources can be distinguished from that of the spontaneous fission source. Such comparisons prevent the spoofing, i.e., intentional imitation, of a fission source by an (α,n) neutron source. This instructional laboratory is designed for nuclear engineering and physics students interested in organic scintillators, neutron sources, and nonproliferation radiation measurement techniques.

In-toto non-destructive assay methodology for the elimination of metallic waste from particle accelerators after melting

Melting of metallic waste reduces the waste volume, allows more accurate radiological characterization, and minimizes handling at the waste production site. This paper proposes a new non-destructive assay methodology to radiologically characterize low- and intermediate-level (LILW) waste before melting. A non-destructive assay technique is developed and qualified using geometry optimization technique and sample analysis after melting. Additionally, we present an operational methodology to predict the activity values of the major gamma emitters based on the average dose rate measurements.

Neutron resonance fission neutron analysis for nondestructive fissile material assay

We propose neutron resonance fission neutron analysis (NRFNA), an active nondestructive assay (NDA) technique, to improve the capability to identify and quantify a small amount of fissile material in a sample. NRFNA uses pulsed neutrons to induce fission reactions in the sample. Fission neutrons are detected by a neutron-gamma pulse shape discrimination (PSD) scintillation detector with time-of-flight (TOF) technique. The obtained nuclide-specific resonance peaks in the neutron energy spectrum provide information to identify and quantify a fissile material in the sample. The possibility of using PSD for NRFNA was confirmed through a test experiment using a natural uranium sample. We successfully observed the resonance peaks from 235U(n,f) reaction and showed that NRFNA would be useful for measuring a small amount of fissile material in a sample.

Analysis of coal ash samples from thermal power plants of India for their gallium content using NAA and EDXRF techniques

Coal fly ash (CFA) is an important secondary source for the recovery of gallium (Ga) which has a high potential for its wide applications in many strategic fields such as cellular communications and direct broadcast satellite.Various coal fly and bottom ash samples obtained from thermal power plants located in different parts of India were investigated for their gallium content using NAA and EDXRF techniques. The concentration of gallium in NIST SRM 1633b CFA is not available in NIST certificate and hence was established using k0 based IM-NAA method along with the other certified elements like As, Ce, Co, Eu, Fe, K etc. In addition, the gallium concentration in the NIST SRM was also obtained by relative NAA using Ga2O3 as a standard towards the validation of IM-NAA. Using the NIST SRM, the gallium content in the ash samples of Bituminous coal collected from South Central and Eastern India was found to be from 17.2 to 47.9 mg/kg whereas the same was 6.3–33.3 mg/kg for the ash samples of Lignite coal collected from the South-Central India and Western India. The gallium concentrations obtained by NAA in coal ash samples were compared with another non-destructive assay technique, i.e. EDXRF and the results are found to be in good agreement. These samples were proved to be a potential secondary source of gallium, available in India.

Improved experimental evaluation and model validation of a 252Cf irradiator for delayed gamma-ray spectrometry applications

We present experimental and MCNP modeled results of neutron flux monitoring in an irradiator, which consists of a252Cf sealed source housed in the center of a cubic block of high-density polyethylene. This work is contributing to our development in the delayed gamma-ray spectroscopy non-destructive assay technique for nuclear safeguards and security applications. We evaluated the thermal and resonance flux values using the Westcott foil activation method for two measurement position count rates of 197Au and 115In. Comparisons between the calculated thermal fluxes from measurements and those from MCNP models show deviations of ~20%. However, both the measured and modeled thermal fluxes are two orders of magnitude higher than the fluxes at the resonance peak energies at both measurement positions, demonstrating that the irradiator is sufficient. Furthermore, this work showed that with a252Cf sealed source of a strength of 4.4 × 106 n.s−1, a thermal flux in the range 2–5 × 104 n.s−1.cm−2 is achievable in our compact irradiator making it suitable for active analysis techniques such as DGS.

Evaluation of high-energy delayed gamma-ray spectra dependence on interrogation timing patterns

The Integrated Support Center for Nuclear Nonproliferation and Nuclear Security of the Japan Atomic Energy Agency is developing delayed gamma-ray spectroscopy (DGS) for nuclear safeguards. Intended to supplement the present verification methods of spent nuclear fuel, we are evaluating the DGS capability to quantify the U and Pu fissile composition using this non-destructive assay technique. Toward this goal, we perform experiments using the PUNITA instrument in collaboration with the European Commission Joint Research Centre in Ispra, Italy. Our studies presented here focus on evaluating the gamma-ray spectral dependence on the fission product decay times. Specifically, we investigated multiple time patterns to evaluate the observable gamma rays useful to quantify the 239Pu and 235U contributions. Of the time patterns tested, spectral distinction is best with a 60-s irradiation followed by a 60-s gamma-ray measurement for a comparable one-hour total interrogation time.

Model design of a deuterium-deuterium neutron generator moderator and evaluation for delayed gamma-ray nondestructive assay for safeguards verification

ABSTRACT The Japan Atomic Energy Agency is evaluating delayed gamma-ray spectroscopy as an active interrogation nondestructive assay technique to quantify the fissile composition of nuclear materials for improved safeguards verification in nuclear fuel cycle facilities. In this paper, we optimize a moderator to thermalize the 2.5 MeV neutrons emitted from a deuterium–deuterium (D–D) neutron generator. The moderator is optimized to maximize the thermal neutron flux, maintain a high moderator quality factor, and be compact for easy installation in the analytical laboratories of reprocessing facilities. With the objective of applying our system to high-radioactivity nuclear materials (e.g. spent fuel solution), we applied this moderator’s flux to an interrogation using delayed gamma-ray spectroscopy for various sample sizes and timing patterns. The results indicate that our 78-cm3 system is sufficiently able to use commercially available D–D neutron generators with neutron emission rates less than 109 n/s for 25- and 100-ml solutions under specific interrogations.

Identification of short-lived fission products in delayed gamma-ray spectra for nuclear material signature verifications

Short-lived Fission Products (FPs) of both uranium (U) and plutonium (Pu) isotopes were identified in neutron irradiated nuclear material (NM) standards in the high-energy gamma-rays range in the perspective of the utilisation of the Delayed Gamma-ray Spectrometry (DGS) technique for nuclear material signature verifications. DGS is a promising active Non-Destructive Assay (NDA) technique that found applications not only in nuclear safeguards but also in nuclear security.This paper presents results of measurement campaigns that include the development of experimental set-up, sample irradiation, gamma-ray spectra acquisitions and qualitative analysis of the obtained gamma-spectra. The experimental setup-up is based on an HPGe planar detector of high-energy resolution and an irradiation chamber made of high-density polyethylene (HDPE) that accommodates a sealed 252Cf source of 50 MBq activity. The delayed gamma-ray spectra of standard samples of U and Pu neutron interrogated samples are successfully acquired using the cyclic neutron irradiation method. 35 and 24 FPs with half-lives ranging from dozens seconds to minutes were identified in the irradiated U and Pu samples, respectively. In the energy range 2–4.9 MeV of the acquired DGS gamma-ray spectra; we have successfully identified 72 gamma-ray peaks of 21 FPs in U and Pu standard samples. This large number of identified FPs is auspicious to the full development of DGS as new active NDA technique for nuclear signature verifications.

Utilizing nuclear data in delayed gamma-ray spectroscopy inverse Monte Carlo analysis

Safeguards verification of uranium and plutonium in high-radioactivity nuclear material is currently performed using destructive analysis techniques. However, the preparation method is a burden on both the safeguards inspectors and facility operators. While nondestructive assay (NDA) techniques would improve the efficiency and time, there are no passive NDA techniques available to directly verify the U and Pu content. As an alternative, the JAEA and JRC are collaboratively developing the Delayed Gamma-ray Spectroscopy (DGS) active-interrogation NDA technique to evaluate the fissile composition from the unique fission product yield distributions. To analyze the data we are developing an Inverse Monte Carlo (IMC) method that simulates the interrogation and evaluates the individual contributions from the mixed nuclear material to the composite spectrum. While the current nuclear data affects the ability to evaluate the composition, the IMC analysis method can be used to determine the systematic uncertainty contributions and has the potential to improve the nuclear data. We will present the current status of the DGS collaborative work as it relates to the development of the DGS IMC analysis.

Ability of non-destructive assay techniques to identify sophisticated material partial defects

This study explores the ability of non-destructive assay techniques to detect a partial material defect in which 100 g of plutonium are diverted from the center of a 1000 g can of PuO2 powder. Four safeguards measurements techniques: neutron multiplicity counting, calorimetry, gravimetry, and gamma ray spectroscopy are used in an attempt to detect the defect. Several materials are added to the partial defect PuO2 can to replicate signatures of the diverted material. 252Cf is used to compensate for the doubles neutron counts, 241Am is used to compensate for the decay heat, and aluminum is used to compensate for the weight. Although, the doubles and triples difference before and after diversion are statistically indistinguishable with the AWCC in fast and thermal mode, the difference in the singles counts are statistically detectable in both modes. The relatively short half-life of 252Cf leads to a decrease (three sigma uncertainty) in the doubles neutron counts after 161 days. Combining this with the precise quantity of 241Am needed (10.7 g) to mimic the heat signature and the extreme precision in 252Cf mass needed to defeat neutron multiplicity measurements gives reassurance in the International Atomic Energy Agency's ability to detect partial material defects.

Investigation of a dynamic measurement methodology for fast detection of gross defects in regularly distributed nuclear material samples

Non-Destructive Assay (NDA) techniques are widely used to verify Nuclear Materials (NM). Most of these techniques are static ones for which, the measuring device and the assayed samples are located at fixed positions during measurements. While assaying regularly distributed NM, attenuation and screening effects may contribute negatively to the accuracy of the results, especially for relatively high density materials. Another factor that may affect the accuracy is the allowable time of measurement. Detection of gross defects in such materials could be achieved more accurately and much faster by employing dynamic measurements. In this work, an investigation for a proposed Non-Destructive Dynamic (NDD) measuring system is presented. The system is assumed to detect gross defects in nuclear fuel assemblies of EK-10 type more accurately and faster than other traditional systems based on static measurements. Different scenarios were considered and studied using the MCNP5 Code. The results showed that the investigated method could be easily applied to detect gross defects in regularly-distributed NM samples.

Model design of a compact delayed gamma-ray moderator system using 252Cf for safeguards verification measurements

Delayed gamma-ray spectroscopy (DGS) is an active-interrogation nondestructive assay technique that can be used to determine the composition of a nuclear material sample by comparing the ratios of fission product gamma-ray peak intensities. However, high-radioactivity nuclear material (HRNM) contains long-lived fission products that can overwhelm a detector and shielding must be used to reduce the count rate, while minimally affecting those gamma rays from short-lived fission products with energy above 3-MeV. To compensate for the signal loss through the shielding, low-energy neutrons are required to induce more fission events from the high thermal cross-section of fissile nuclides. To improve practical safeguards DGS capabilities, we are developing a compact interrogation system to moderate ∼ 2-MeV neutrons that are easier to moderate than 14-MeV neutrons from standard deuterium-tritium generators. This work describes the optimization of an ideal moderator system for a 252Cf neutron point source that results in a neutron fluence of 25.9×10−4 cm−2nsource−1 passing through the sample space with ≳ 70% of those below 1-eV. Modifications for practical fabrication resulted in ⩽20% reductions of the flux compared to the optimized ideal design. Finally, evaluations made of HRNM DGS signals and backgrounds conclude that a 252Cf source intensity of 8.9×107 neutrons per second is required for a single-pass interrogation within this ideal moderator system. However, this can be as low as 3.6×106 neutrons per second using smaller samples that require less shielding.

Development of Neutron Resonance Transmission Analysis as a Non-Destructive Assay Technique for Nuclear Nonproliferation<sup> </sup>

Development of Neutron Resonance Transmission Analysis as a Non-Destructive Assay Technique for Nuclear Nonproliferation<sup> </sup>

Applicability study of the photofission based nuclear material isotopic composition measurement method on the Thorium-Uranium system

A principle of new nondestructive assay (NDA) technique based on the photofission reaction ratio (PFRR) has been developed, which is aimed at measuring the isotopic composition of nuclear fuel materials without relying on their self-generated neutron information. The feasibility of the PFRR method was validated in the 235U-238U system previously. However, applicability of the PFRR method in the other material is not yet validated. In this study, applicability of PFRR method to the 232Th-233U system was confirmed, PFRR method showed good reproducibility of predicted value of 232Th and 233U isotopic composition.

Numerical convergence and validation of the DIMP inverse particle transport model

Abstract The data integration with modeled predictions (DIMP) model is a promising inverse radiation transport method for solving the special nuclear material (SNM) holdup problem. Unlike previous methods, DIMP is a completely passive nondestructive assay technique that requires no initial assumptions regarding the source distribution or active measurement time. DIMP predicts the most probable source location and distribution through Bayesian inference and quasi-Newtonian optimization of predicted detector responses (using the adjoint transport solution) with measured responses. DIMP performs well with forward hemispherical collimation and unshielded measurements, but several considerations are required when using narrow-view collimated detectors. DIMP converged well to the correct source distribution as the number of synthetic responses increased. DIMP also performed well for the first experimental validation exercise after applying a collimation factor, and sufficiently reducing the source search volume's extent to prevent the optimizer from getting stuck in local minima. DIMP's simple point detector response function (DRF) is being improved to address coplanar false positive/negative responses, and an angular DRF is being considered for integration with the next version of DIMP to account for highly collimated responses. Overall, DIMP shows promise for solving the SNM holdup inverse problem, especially once an improved optimization algorithm is implemented.

Comparative analysis of nondestructive ASSAY techniques for 235U and 239Pu in structural materials at a high gamma background level

Potential techniques to identify small quantities of fissionable materials (FM) (0.001% wt.) in conditions of a high gamma background level have been reviewed and compared, and the optimal possibility for the nuclear material (NM) control in spent fuel assemblies (SFA) has been selected. It was found through numerical simulation that a system based on a passive neutron control method could be used to detect FMs indirectly when the spent nuclear fuel (SNF) burn-up and cooling time are known. Two types of detectors have been compared: 3He counters and 235U-based fission chambers. Better application prospects of 3He counters, based on the SNM-18 neutron counter, have been shown and drawbacks of passive control technique have been pointed out.An active neutron control has been found to be the best way to address the problem considered. The system's computational model shows that the signal exceeds the triple background error (both for ambient and intrinsic background from Cm isotopes) more than twelvefold. To improve the signal recording efficiency, the system has been modified to allow for irregularities in the geometrical position of structural materials (SM) in the measuring chamber.The proposed procedure makes it possible to determine in a short time the content of 239Pu, 242Cm, and 244Cm in an SFA. After the quantity of 239Pu is determined, it is possible to estimate the content of other isotopes (Am, U, Np) due to the constancy of the 239Pu mass ratio to the mass of the actinide identified.

Comparative analysis of non-destructive methods to control fissile materials in large-size containers

The analysis of various non-destructive methods to control fissile materials (FM) in large-size containers filled with radioactive waste (RAW) has been carried out. The difficulty of applying passive gamma-neutron monitoring FM in large containers filled with concreted RAW is shown. Selection of an active non-destructive assay technique depends on the container contents; and in case of a concrete or iron matrix with very low activity and low activity RAW the neutron radiation method appears to be more preferable as compared with the photonuclear one.

Delayed gamma-ray spectroscopy combined with active neutron interrogation for nuclear security and safeguards

For the purpose of nuclear security and safeguards, an active neutron interrogation non-destructive assay technique, Delayed Gamma-ray Spectroscopy (DGS), is under development. The technique of DGS uses the detection of decay γ rays from fission products to determine ratios of fissile nuclides in a sample. A proper evaluation of such γ-ray spectra requires integration of nuclear data such as fission cross-sections, fission yields, half-lives, decay-chain patterns, and decay γ-ray yields. Preliminary DGS experiments with the Pulsed Neutron Interrogation Test Assembly, named PUNITA, of the European Commissions’ Joint Research Center have been performed. Signals of delayed γ ray from nuclear materials were successfully observed.