AmLi Sources Research Articles

Study on the Fast Neutron Multiplicity Measurement of Uranium Material

As a non-destructive testing and analysis technology, the neutron multiplicity measurement method plays an important role in the field of arms control verification. Great progress has been achieved in the fast neutron multiplicity measurement of plutonium, while there are few studies that have been conducted in the fast neutron multiplicity measurement of uranium. In this study, a set of fast neutron multiplicity measurement devices based on the BC501A liquid scintillation detector was built, and two small mass uranium samples with low density were studied. It is found that the change of the total neutron counting rate is irregular, and the coincidence counting rate will increase linearly with the mass. It is due to the energy loss caused by the scatter of the Am-Li source neutron with the sample, resulting in part of the neutron energy below the threshold. A simulation detection system is built by Geant4 to verify the reliability. On this basis, four kinds of packaging materials such as steel, copper, aluminum, and graphite were studied. The measurement results of different materials were compared and analyzed. The thickness of the material was studied, and the functional relationship curve was fitted. In this study, the fast neutron multiplicity measurement of small-mass and low-density uranium samples was carried out and verified. The influence of different packaging materials on the multiplicity measurement was analyzed, which is of great significance for the development of the fast neutron multiplicity measurement technology.

Gamma-blindness & neutron detection efficiency assessments with centrifugally tensioned metastable fluid detectors in low-scatter environment

A range of tests and associated analyses were performed in a state-of-art low scatter environment to demonstrate and validate the performance of the (Medium) Centrifugally Tensioned Metastable Fluid Detector (MCTMFD) for gamma blind detection of fast and epithermal neutrons for a variety of configurations, using Cf-252 and Am-Li isotope neutron sources. All measurements were performed in the Low Scatter Room (LSR) of Radiological Exposures & Metrology Laboratory at Pacific Northwest National Laboratory (PNNL). Experiments were performed to evaluate intrinsic neutron detection efficiency, ɛint n, the intrinsic gamma-neutron detection efficiency, ɛintγn, and the gamma absolute rejection ratio for neutrons (GARRn) as a function of tensioned metastable state negative pressure, Pneg, varying from 2.5 bar towards 9 bar. The specific embodiment of the MCTMFD utilized for the experiments at PNNL included a 16 cm 3 sensitive volume configured for either fast neutron detection using Decafluoropentane (DFP) as the detection medium, or fast and thermal neutron detection (using DFP with a chemical mixture composed of natural boron). Experiments were performed with Cf-252 neutron sources ranging in intensity from 1 × 104 n/s to 3.45 × 105 n/s, an Am-Li neutron source with intensity of 2.5 × 104 n/s, used in both a bare configuration and in combination with a 30 cm OD moderating sphere of D2O, with and without a 0.5 mm Cd shell, and intense Cs-137 and Co-60 gamma sources offering dose rates ranging from 27 mR/h to 53 R/h. It was found that the MCTMFDs remained gamma-blind (ɛintγn < 10−11) even under 53 R/h gamma background; the associated intrinsic fast neutron detection efficiency (ɛint n) varied from ∼0.2% for Pneg ∼2.5 bar, ∼10% for Pneg ∼5.5 bar and, ∼30% for Pneg ∼9 bar for a Cf-252 fission spectrum. In comparison, for the softer (Am-Li) spectrum, the ɛint n varied from ∼1% for Pneg ∼5.5 bar and approached the ɛint n values for Pneg states above 7 bar. The same MCTMFD when using a borated detection fluid with 20% B-10, offered a ×10–20 improvement in ɛint n for Pneg states ranging from 4.5–5 bar, for detecting epithermal neutrons from the Am-Li source (placed within a 30 cm OD, D2O moderator with and without a 0.5 mm Cd shield).

Configuration and impurity quantification of AmLi sources using radiography and gamma spectroscopy

This paper presents the non-destructive assay of two AmLi sources used for detector characterization and active interrogation at PNNL, identified as MRC-101 and MRC-103. First, a detailed description of the internal configuration of 2724-BT encapsulated MRC sources is established. X-ray radiographs of the AmLi sources are reported for the first time to verify the dimensions and orientation of the encapsulation. Notably, a density variation was observed as a contrast change across the height of the source cavities. Contrary to commonly made assumptions in literature, it is indirectly observed that the AmLi source/target material only occupies a portion of the source cavity. The intensity of the prompt gamma-rays resulting from the inelastic scatter of alpha particles on 7Li was used to calculate the mass ratios of LiH:AmO2. The gamma-ray spectra revealed the presence of numerous impurities in both source and target material (243Am, 237Np, 154Eu, 23Na, and 9Be). Neutron-gamma production ratios were applied to the intensity of the Doppler-broadened peaks from alpha-induced reactions with 7Li, 9Be, and 23Na to calculate neutron rates for both sources. A neutron/alpha production yield (derived in this work) was then related to the atomic concentration of these isotopes to find the approximate masses within each source. Ultimately, the AmLi source density was estimated with prompt gamma analysis based on partial source volumes as 0.45±0.08g∕cm3.

Genetic algorithm optimization of tin–copper graded shielding for improved plutonium safeguards measurements

Plutonium nuclear material safeguards is essential to ensure that states do not divert plutonium to clandestine nuclear weapons programs. Traditional safeguards methods rely on correlated neutron measurements. Detection systems based on organic scintillators, which are sensitive to fast neutrons and photons, are gaining support for use in safeguards applications due to increased capabilities and shorter measurement times compared to the current state-of-the-art. A drawback of organic scintillators in plutonium measurements is data throughput limitations; 241Am builds up in plutonium samples and emits 60-keV gamma rays with a high specific activity. In this work, a genetic algorithm is used to design graded shielding made of tin and copper to attenuate the 60-keV gamma rays. The shield is designed to limit the attenuation of the higher-energy, fission photons, because fission photons can be used as a signature. The genetic algorithm optimization takes days to run, while a complete search space exploration would take years. The genetic algorithm was run for different levels of 60-keV gamma-ray attenuation, and the thickness of the required shielding layers was determined at each level. Tin–copper shielding was placed in front of an array of organic scintillation detectors for a measurement of approximately 4.5 kg of alpha-phase, weapons-grade plutonium, an AmLi source, and a 252Cf source. We show that the shield successfully attenuated low-energy, non-fission gamma rays, which significantly improved the organic scintillator performance.

Characterization of two extraordinary AmLi neutron sources

Two legacy AmLi sources have been actively used across several projects within Pacific Northwest National Laboratory, however, applicability of source measurement data has been substantially limited due to the lack of original manufacturer documentation and cross comparison to other sources. This paper summarizes measurements conducted to better quantify source radionuclide content, neutron energy spectrum and emission rate of the AmLi sources. The measured neutron energy spectra of AmLi sources produced by Monsanto Research Corporation have not been found in open literature before and therefore the focus of the study was to obtain the high definition energy distribution of the sources. The initial gamma spectrometry measurements, aimed to estimate the radionuclide content and potential impurities, revealed a significant contamination of the source material with beryllium. Consequent neutron spectrum measurements confirmed beryllium contamination through the presence of a high energy tail contributing up to 17% to the total neutron emission rate. The obtained lithium-only portion of the measured source energy distribution was compared against other AmLi spectra discussed in the past and recent literature. Determination of the total neutron yields of the sources was made with measurements in a neutron well counter. Detection efficiency of the well counter was calculated specifically for these sources via detailed Monte Carlo simulations based on the obtained neutron energy distribution.

Variations in AmLi source spectra and their estimation utilizing the 5 Ring Multiplicity Counter

Active-mode assay systems are widely used for the safeguards of uranium items to verify compliance with the Non-Proliferation Treaty. Systems such as the Active-Well Coincidence Counter (AWCC) and the Uranium Neutron Coincidence Collar (UNCL) use americium-lithium (AmLi) neutron sources to induce fissions which are measured to determine the sample mass. These systems have historically relied on calibrations derived from well-defined standards. Recently, restricted access to standards or more difficult measurements have resulted in a reliance on modeling and simulation for the calibration of systems, which introduces potential simulation biases. The AmLi source energy spectra commonly used in the safeguards community do not accurately represent measurement results and the spectrum uncertainty can represent a large contribution to the total modeling uncertainty in active-mode systems.The 5-Ring Multiplicity Counter (5RMC) has been used to measure 17 AmLi sources. The measurements showed a significant spectral variation between different sources. Utilization of a spectrum that is specific to an individual source or a series of sources will give improved results over historical general spectra when modeling AmLi sources. Candidate AmLi neutron spectra were calculated in MCNP and SOURCES4C for a range of physical AmLi characteristics. The measurement and simulation data were used to fit reliable and accurate AmLi spectra for use in the simulation of active-mode systems. Spectra were created for average Gammatron C, Gammatron N, and MRC series sources, and for individual sources. The systematic uncertainty introduced by physical aspects of the AmLi source were characterized through simulations. The accuracy of spectra from the literature was compared.

Monitoring of 235U content of fuel assemblies for nuclear reactors based on neutron and γ-ray scintillation detector systems

A model of the setup for determining the 235U content of a fuel assembly for WWER-440 reactors is described. The setup operates based on the digital technology for discriminating neutron and photon scintillation responses. An AmLi source is used as an external neutron source. A three-channel system for detecting fission particle coincidences is based on EJ-309 scintillators. A set of digital codes for obtaining and analyzing neutron-neutron (nn), neutron-γ (nγ), and γ-γ (γγ) coincidences has been developed. Dependences of double coincidences have been experimentally obtained and analyzed with the aim of determining the 235U content of the fuel assembly. It is shown that fission neutrons and coincidences of the nn and nn + γn + nγ types are the best informative parameters for the 235U content of the fuel assembly.

Monte carlo feasibility study of an active neutron assay technique for full-volume UF6 cylinder assay using a correlated interrogation source

Uranium cylinder assay plays an important role in the nuclear material accounting at gas centrifuge enrichment plants. The Passive Neutron Enrichment Meter (PNEM) was designed to determine uranium mass and enrichment in 30B and 48Y cylinders using total neutron and coincidence counting in the passive mode. 30B and 48Y cylinders are used to hold bulk UF6 feed, product, and tails at enrichment plants. In this paper, we report the results of a Monte-Carlo-based feasibility study for an active uranium cylinder assay system based on the PNEM design. There are many advantages of the active technique such as a shortened count time and a more direct measure of 235U content. The active system is based on a modified PNEM design and uses a 252Cf source as the correlated, active interrogation source. We show through comparison with a random AmLi source of equal strength how the use of a correlated driver significantly boosts the active signal and reduces the statistical uncertainty. We also discuss ways in which an active uranium cylinder assay system can be optimized to minimize background from 238U fast-neutron induced fission and direct counts from the interrogation source.

Determining the “doubles” count rate for AmLi sources

A simple calculation model for AmLi neutron sources is considered: the primary neutrons from Li ( α , n ) reaction are assumed to be randomly emitted from the centre of a spherical source, and to induce the time-correlated secondary neutrons via fission process. The effect of Am 241 (n, fission) was calculated and found to be very small. However, it largely contributes to the “doubles” count rate of AmLi determined by neutron counting technique.

Non-destructive actinide assay of neutron sources

Transuranic neutron sources were assayed using γ-spectrometry and neutron counting, with respect to revealing illicit trafficking of radioactive and nuclear materials. Source type and constituents were identified by γ-spectrometry. For determining the source strength of 244Cm, 252Cf, AmBe, AmLi and 238PuBe sources, a passive neutron coincidence collar was used with 3He counters in two moderator configurations. The electronics consist of independent channels of pulse amplifiers and discriminators as well as a shift register for coincidence counting. From the neutron output of the sources measured by gross neutron counting, the actinide content (or activity) can be assessed by adopting specific spontaneous fission and (α,n) reaction yields of individual isotopes from the literature. Various types of the sources can be characterized by the ratio of the real coincidence counts to total (singles) counts, R/T. This quantity is a specific constant for different types of spontaneously fissioning sources, not depending on their strength, while exhibits a power function increasing with the strength of (α,n) sources. An individual source of the latter type can additionally be characterized by the ratio of R/T to the product of detection efficiency and a factor taking into account coincidence loss, irrespective of the instrumental details (detector type and size, moderator thickness, etc.).

Quantitative NDA of isotopic neutron sources

A non-destructive method for assaying transuranic neutron sources was developed, using a combination of γ-spectrometry and neutron correlation technique. Source strength or actinide content of a number of PuBe, AmBe, AmLi, 244Cm, and 252Cf sources was assessed, both as a safety issue and with respect to combating illicit trafficking. A passive neutron coincidence collar was designed with 3He counters embedded in a polyethylene moderator (lined with Cd) surrounding the sources to be measured. The electronics consist of independent channels of pulse amplifiers and discriminators as well as a shift register for coincidence counting. The neutron output of the sources was determined by gross neutron counting, and the actinide content was found out by adopting specific spontaneous fission and ( α , n ) reaction yields of individual isotopes from the literature. Identification of an unknown source type and constituents can be made by γ-spectrometry. The coincidences are due to spontaneous fission in the case of Cm and Cf sources, while they are mostly due to neutron-induced fission of the Pu isotopes (i.e. self-multiplication) and the 9Be(n,2n) 8Be reaction in Be-containing sources. Recording coincidence rate offers a potential for calibration, exploiting a correlation between the Pu amount and the coincidence-to-total ratio. The method and the equipment were tested in an in-field demonstration exercise, with participation of national public authorities and foreign observers. Seizure of the illicit transport of a PuBe source was simulated in the exercise, and the Pu content of the source was determined. It is expected that the method could be used for identification and assay of illicit, found, or not documented neutron sources.

Fast neutron dosimetry with CR-39 using electrochemical etching

A Monte Carlo method is used in order to calculate the response (tracks⋯cm −2 ⋯ μSv −1 ) of a fast neutron dosemeter, composed of a polyethylene radiator and CR-39 plastic track detector, to monoenergetic neutron beams at normal incidence and taking into account the electrochemical etching conditions. The method is applied to neutrons having energies in the range 0.08–10 MeV. The total number of protons emerging from the radiator is calculated, as well as their energetic and angular distribution. Detection efficiency as a function of proton energy and angle is taken into account. The theoretical calculation is compared with experimental results obtained for Am-Li and Pu-Be sources for protonic equilibrium radiator thicknesses