Neutron Activation Research Articles

Post irradiation examinations of FeCrAl cladding in PWR conditions

FeCrAl alloys have been one of the prominent Accident Tolerant Fuel (ATF) cladding material candidates, primarily due to their excellent oxidation resistance in high-temperature steam conditions when compared to Zr-based alloys. Prototypic irradiation of fueled FeCrAl rods is a fundamental step in confirming the integral performance behavior during in-pile conditions. Early generation C26M cladding, fabricated through wrought metallurgy techniques, was fueled with UO2 fuel pellets and irradiated in a pressurized water loop in the Advanced Test Reactor (ATR) at the Idaho National Lab (INL) to a burnup (BU) of ∼25 GWd/tHM. After irradiation, the rodlets were nondestructively and destructively examined. Nondestructive examinations included visual exams, profilometry, and gamma scanning. These examinations highlight the unique deposits, BU profile of the rodlets as well as the migration path of fission gases within the rodlet, and typical diametrical morphology of fuel rodlets. During handling, brittle failure of one end cap on one pin occurred. Destructive examinations included microscopy and mechanical testing. Radial cross sections of the cladding were analyzed metallographically through light optical microscopy (LOM), scanning electron microscopy (SEM) highlighting a unique corrosion morphology and micro-cracking (∼20–30 μm past the main oxide layer) at the metal-oxide interface. Ring compression testing (RCT) was used to elucidate the mechanical property change of the FeCrAl cladding after neutron irradiation. In contrast to the non-irradiated material, which remained ductile at all test temperatures between 25 °C and 250 °C, the irradiated cladding fractured in a brittle manner at 50 °C and below. The results of the tests show that some challenges remain in the development of FeCrAl cladding for LWR cladding applications including improvement of in-reactor waterside corrosion performance and the retention of ductility after neutron irradiation.

Field Test of the MiniRadMeter Gamma and Neutron Detector for the EU Project CLEANDEM.

In the framework of the H2020 CLEANDEM project, a small robotic vehicle was equipped with a series of different sensors that were developed for the preliminary inspection of areas possibly contaminated by radiation. Such unmanned inspection allows to identify dangerous locations prior to the possible start of human operations. One of the developed devices, named the MiniRadMeter, is a compact low-cost sensor that performs gamma and neutron radiation field mapping in the environment. The MiniRadMeter was successfully tested in a simulated field mission with four "hidden" radioactive sources and a neutron generator. In this work, we describe the test procedure and the results, which were supported by the outcome of dedicated Monte Carlo simulations.

Design of a compact and portable space neutron spectrometer based on Monte Carlo

With the rapid development of space exploration, the detection of space neutron radiation is becoming increasingly important. The currently widely used Bonner sphere spectrometer have drawbacks such as large size and weight, as well as low fault tolerance, when detecting space neutron spectra. This paper describes in detail a new type of space neutron spectrometer (SNS), which has two different specifications to adapt to the directional and non-directional neutron field environment, and can measure the directional neutron energy spectrum. For the directed neutron field, SNS integrates 12 3He thermal neutron counters (diameter 3 cm: 3, diameter 4 cm: 6, diameter 5 cm: 3) and uses cylindrical polyethylene as a moderator. For non-directed neutron fields, SNS integrates 9 3He thermal neutron counters (diameter 3 cm: 4, diameter 4 cm: 3, diameter 5 cm: 2) located in a single structure made of polyethylene, boron-containing polyethylene and gadolinium. The device is capable of providing a strong directional response in the energy range of thermal neutrons up to 20 MeV, with little sensitivity to neutrons coming from directions other than the axis of the cylinder. The Monte Carlo transport code FLUKA was used to determine the final configuration of the instrument, including the arrangement, number, and position of thermal neutron counters. In addition, the response matrix of the instrument was calculated using FLUKA code. This device can replace traditional Bonner sphere spectrometer for measuring space neutrons, and it also provides reference value for downsized and lightweight neutron spectrometers on the ground.

Neutron activation analysis for development of organic coffee leaves reference material

Neutron activation analysis for development of organic coffee leaves reference material

Measurement of the 117mSn, 125mTe, 135mBa and 195mPt half-lives for reactor dosimetry application

Nuclear isomers are investigated to perform precise epithermal neutron dosimetry. One key physical property for reactor dosimetry is the precise knowledge of the isomer half-life. In the list of potential candidates, the half-life values available in the literature or in the recommended database for reactor dosimetry for 117mSn, 125mTe, 135mBa and 195mPt show some discrepancy. New half-life measurements are presented in this work. Those isomers were produced by neutron activation at the JSI TRIGA Reactor from tin, tellurium, barium (all three of them enriched in their parent of interest) and natural platinum dosimeters. Precise half-lives measurements were performed at the MADERE platform using gamma and X-ray spectrometry. Special care has been taken on the uncertainty determinations.

A review of criticality dosimetry at the Y-12 National Security Complex and practical importance of dose accuracy in emergency response

A nuclear criticality results in the emission of both neutron and gamma radiation and can produce doses to personnel near the event that exceed 0.1 Gy (10 rad). The primary purpose of nuclear accident dosimetry is to rapidly identify affected personnel in need of prompt medical treatment and to reassure personnel who have been only minimally exposed. While accurate dosimetry is desired, it must be recognized that dose determinations made from whole-body dosimeters or simple triage methods are very rough estimates and contain significant uncertainties. Even when accounting for factors like varying neutron energy spectra, mean photon energies, body orientation within the radiation field, and transient effects on dosimeter response, etc., the end value is a dosimetric quantity defined for very specific radiological conditions and determined within a simple phantom usually at a single depth. Of more importance is the biological response to the radiation, which will vary by person and can be affected by the individual's radiation sensitivity, age, gender, mass, and underlying health conditions. The overall biological, person-specific response to a given dose cannot be precisely determined except by patient symptom observation and individual biological dosimetry (e.g. chromosome analysis, lymphocyte ratios, etc.). This work describes and discusses the criticality accident dosimetry program at the Y-12 National Security Complex, a United States Department of Energy National Nuclear Security Administration facility. The primary goals of the Y-12 accident dosimetry program are, among others, the rapid identification of significantly exposed persons, prompt routing of exposed workers for medical evaluation and treatment, and the ultimate processing of dosimeters to assign doses to personnel.

On the iron content of Mn-Ni-Si-rich clusters that form in reactor pressure vessel steels during exposure to neutron irradiation

Solute cluster formation during service is the primary cause of embrittlement in reactor pressure vessel (RPV) steels. Atom probe tomography (APT) has consistently shown that the solute clusters are enriched in Cu, P, Mn, Ni, and Si compared to the matrix. However, there is pronounced disagreement within the literature as to the Fe content of the solute clusters in low-Cu RPV steels; some authors report Fe contents in excess of 80 at.% whilst others attribute all measured Fe in the clusters to aberrations from APT and report Fe-free features. This discrepancy has profound implications for determining whether cluster formation is radiation-induced or radiation-enhanced.In this article, we perform a systematic analysis of the published literature to demonstrate the variance present in the measured Fe content of solute clusters characterised by APT. We investigate potential explanatory variables for this variation and discuss the difficulties in performing statistical analyses on these data. To determine the impact of APT aberrations on the measured Fe content of solute clusters, we combine results from: simulations that account for the limited spatial precision in APT; models that predict ion trajectories in the presence of non-uniform electrostatic fields; and experimental data.Our results show that APT aberrations can introduce large amounts of artificial Fe into solute clusters, especially at the small sizes (radius <1.5 nm) typically observed in RPV steels. We also show, however, that some of the variance in Fe content in the literature may be explained by real differences in the clusters’ Fe contents.

Characterization of neutron spectrum of polyethylene-moderated AmBe neutron source using a passive single-cylindrical neutron spectrometer

The changes in the neutron spectrum produced by the AmBe neutron source moderated in the polyethylene sphere need to be characterized. A passive single-moderator neutron spectrometer with indium foil activation neutron detector has been developed to characterize the neutron spectrum of an AmBe source inside a polyethylene sphere. The detector response function was calculated using the MCNPX 2.7 program with IRDFF-II nuclear data library. Measurements were conducted by placing the Single-Cylindrical Neutron Spectrometer (SCNS) on the outer surface of the sphere for 5 h. The 116mIn activity due to neutron activation in each indium foil was measured using a gamma spectrometer with an HPGe detector. By identifying the count value at the peak energy of 1294 KeV and considering an HPGe detector efficiency of 3.2% at the foil position 1 mm above the detector surface, the activity of 116mIn was obtained. The activity value of 116mIn from each indium foil was compared with the MCNPX simulation results. The neutron spectrum was unfolded using the UMG 3.3 program with activity data input of 116mIn for each foil and detector response. A neutron spectrum was obtained with a total neutron flux of 634 ± 60 n/cm2·s, consisting of 25% thermal neutrons, 16% epithermal neutrons, and 61% fast neutrons. When compared with the simulation results, the total neutron flux in the spectrum produced by SCNS-In showed only a small difference of 1%. Based on these neutron spectrum measurements, it was determined that placing the AmBe neutron source inside a 15″ diameter PE-sphere will reduce the fast neutron flux by 78%.

Volume fraction detection in multiphase systems using neutron activation analysis and artificial neural network

This study presents an application of an Artificial Neural Network (ANN) to detect fluids in an annular flow regime using Prompt-Gamma Neutron Activation Analysis (PGNAA). The ANN was trained using gamma-ray spectra resulting from neutron interactions with chemical elements found in fluids typical of multiphase flow in oil exploration. These spectra were generated through mathematical simulation using the MCNP6 Monte Carlo computer code to model nuclear particle transport. A241Am-Be polyenergetic neutron source was simulated for these calculations. Several combinations of fluid fractions were developed to create a dataset used for both training and evaluation of the ANN. The ANN demonstrated robust generalization capabilities by accurately predicting the volume fraction of the three investigated fluids (saltwater, oil, and gas), even for cases not included in the training phase. The combination of ANN and PGNAA proved effective for analyzing multiphase systems, with over 92% of all showing errors of less than 5%.

Stress and temperature dependence of irradiation creep in zircaloy-4 studied using proton irradiation

This paper describes a series of experiments conducted to investigate the in-situ irradiation creep behavior of Zircaloy-4 using proton irradiations as a surrogate to neutrons. The first series of experiments investigated the impact of the initial irradiation-induced defect evolution during the 0 – 0.1 dpa regime on the subsequent in-situ steady-state behavior derived from experimentation. These experiments were conducted at a constant applied load of 85 MPa, a constant damage rate of 1.58 × 10−6 dpa/s, and were repeated at both 250 °C and 350 °C. We also examined the use of a ‘conditioning-irradiation’ step prior to the creep tests on the results derived from subsequent in-situ proton irradiation creep experiments. By extension, we aimed to further develop and refine optimum testing procedures when using proton irradiations to investigate the in-situ creep behavior of nuclear materials. The second series of in-situ proton irradiation experiments were conducted on two Recrystallized (RX) Zircaloy-4 samples in order to investigate the temperature and stress dependence of irradiation creep. One sample was held at a constant load while the temperature was varied in the range of 275 – 350 °C, and the other was held at a constant temperature while stress was varied in the range of 65 – 105 MPa. The associated strains and creep rates were measured at each interval and used to determine an activation temperature of 5000 ± 1700 K and a stress sensitivity exponent of 4.3 ± 0.8 for RX Zr-4 over the given temperature and stress ranges. A discussion of potential deformation mechanisms based on competition between bulk diffusion through the lattice and point defect diffusion enabling dislocation climb and glide is given: the relatively high stress dependence suggested the latter is more likely however further investigations will be required to improve the mechanistic understanding. The results presented in this manuscript, including activation temperature, stress sensitivity, and associated creep rates, determined through proton irradiation investigations are closely comparable to those determined from neutron irradiation experiments found in the literature for similar zirconium alloys at similar temperatures and stresses. Coupled with the primary-secondary creep behavior investigations presented, this demonstrates the usefulness of this approach to estimate neutron-irradiation equivalent creep behavior using in-situ proton irradiation.

Radiochemical separation of 161 Tb from neutron irradiated Gd target by liquid-liquid extraction technique

Abstract No-carrier-added (NCA) 161 Tb, which has advisable nuclear properties to be applied for cancer radiotherapy was produced at the Egyptian Second Research Reactor (ETRR-2) by neutron irradiation of natural gadolinium target via indirect nuclear reaction. The radiochemical separation of 161 Tb from irradiated gadolinium target was investigated based on solvent extraction technique using Cyanex 302. Several separation parameters were checked and optimized. According to the obtained results, the separation process of the investigated radioisotopes proceeds in two steps. The first step is an extraction of all of them into the organic phase in which the extraction % (98 %) was optimized at pH 4, 0.15 M of Cyanex 302 and 2.5 h extraction time. Moreover, the slope analysis method confirmed the participation of 2 mol of the organic extractant for the separation of 161 Tb from irradiated gadolinium. The second step is the separation of the 161 Tb isotope that was purified by the stripping of 159Gd with a citrate solution at pH 9, which is considered as a highly efficient and promising method for separation and purification of the two radioisotopes.

Assessment of potential toxic elements in soils, sediments, and vegetation in the surroundings of Anapa, Russia.

The study presented here reports the concentration of major, trace, and rare earth elements in soil, sediments, and vegetation samples collected from 13 locations around Anapa City located on the northern coast of the Black Sea in Russia. The neutron activation analysis technique has been used to fulfill this objective. Along with this, the bioconcentration and translocation factors were calculated. Overall, the content of 31 elements was detected in soil and sediments while 20 elements were determined in three types of vegetation: macroalgae (Cystoseira sp. and Ulva sp.), aquatic plants (Phragmites australis), and sea grass (Zostera sp.). The quantified concentration followed the order soil > sediment > vegetation. The phytotoxic levels for Zn, V, Mn, and Fe have been quantified as the highest. Bromine was the most abundant and accumulated in Phragmites australis. Based on the results obtained from this investigation, there is a possibility of contamination in the study area.

Assessing Radiation Fallout in Public Zones near the Nevada National Security Site (NNSS): A Recent Study.

A comprehensive radiological study was conducted in the surrounding public zones of the Nevada National Security Site to identify traces of resuspended radioactivity and heavy elemental contamination that might have resulted from various activities. The study used passive and active nuclear methods, specifically gamma spectrometry and instrumental neutron activation analysis, respectively. Passive gamma spectra analysis of air filter papers from various Community Environmental Monitoring Program stations conclusively verified the presence of radionuclides exclusively originating from the natural decay series of 238U and 232Th. Furthermore, gamma spectrometry and instrumental neutron activation analysis of plant samples from surrounding areas of the Nevada National Security Site revealed the absence of any unusual elemental contamination in the environment. These results demonstrated that there was no measurable radiological impact on the public zones surrounding the site resulting from the spread of radioactive materials or toxic heavy metals associated with previous or ongoing activities at Nevada National Security Site. Therefore, the safety of public zones concerning retained radioactivity and harmful elemental contamination arising from Nevada National Security Site operations is negligible. The significance of this study is further pronounced in the current geopolitical context, as it establishes the baseline elemental composition for various desert plants for future reference.

Measuring SET Pulse Widths in pMOSFETs and nMOSFETs Separately by Heavy Ion and Neutron Irradiation

Measuring SET Pulse Widths in pMOSFETs and nMOSFETs Separately by Heavy Ion and Neutron Irradiation

Performance of Iron Phosphate Glass Containing Various Heavy Metal Oxides for Particulate Nuclear Radiation Shielding.

Employees may be exposed to different kinds of ionizing radiation at work. When ionizing radiation interacts with human cells, it can cause damage to the cells and genetic material. Therefore, one of the scientists' primary objectives has always been to create the best radiation-shielding materials. Glass could offer promising shielding material resulting from the high flexibility of composition, simplicity of production, and good thermal stability. The melt-quenching technique was used to create a glass having the following formula: 50% P2O5+20% Na2O+20% Fe2O3+10% X, where X = As2O3, SrO, BaO, CdO, and Sb2O3 mol %. The impact of the different heavy metal additions on the structure of the glass networks was studied using FTIR spectroscopy. Glass's ability to attenuate neutrons and/or charged particles has been theoretically investigated. The performance of the developed glass as a shield was examined by a comparison against commercial glass (RS 253 G18), ordinary concrete (OC), and water (H2O). For charged particle radiations (Electrons, Protons, and Alpha), the shielding parameters like the mass stopping power, the projected range, and the effective atomic number were evaluated, where S5/Sb glass achieves the best performance. In the case of Neutrons, the results values reveal that S3/Ba glass ( ΣR = 0.105) is the best-modified glass for neutron shielding. Among all the investigated glasses, S5/Sb glass composition has a smaller range and provides superior protection against charged particles. In contrast, the S3/Ba glass composition is a superior choice for shielding against neutron radiation.

Synthesis, photoluminescence, Judd-Ofelt parameters, and high energy neutron/charged particle transmission efficiencies of Nd3+ ion-activated sodium-borate glasses

Sodium borate glasses doped with Nd3+ ions and having the molar formulations 75B2O3 – 15Na2O – 9.5x – 0.5Nd2O3, where x = Bi2O3; SrO and Li2O were prepared by employing the melt quenching technique. The glasses were coded as BiNd, SrNd, and LiNd for x = Bi2O3; SrO and Li2O, respectively. The prepared glasses were characterized for their physical and optical attributes. The influence of the glass composition on the photoluminescence and fast neutron (FN) and charged radiation (CR) transmission efficiencies were also investigated. The density of BiNd, SrNd, and LiNd is 3.32, 2.43, and 3.32 g/cm3, respectively. Optical constants of the glasses showed variations with glass composition. The FTIR study of the glass also revealed the structure of the glasses. The BiNd glass has the most intense PL of all the samples, followed by SrNd glass, while LiNd had the weakest PL emission. The value of for BiNd, SiNd, and LiNd is 0.1029 cm-1, 0.1009 cm-1, and 0.0987 cm-1, respectively. The projected ranges of electrons, protons, helium ions, and carbon ions were generally lower for denser XNd glass. The XNd glasses are potentially useful of optical, shielding and dosimetry applications.

Microstructure evolution mechanism of WB-doped Fe-based amorphous composite coating under proton beam irradiation

Fe-based amorphous coatings exhibit exceptional irradiation resistance attributed to their distinct topologically disordered structure, rendering them highly attractive for advanced nuclear energy applications. The incorporation of WB secondary phase doping can notably alter the coating to enhance its operational safety. In this investigation, three different Fe-based composite coatings, with varying WB doping levels of 5 %, 10 %, and 15 % were fabricated through the High-Velocity Oxy-Fuel (HVOF) spraying technique. Irradiation tests were conducted at room temperature utilizing a proton beam with an energy of 1.52 MeV to simulate neutron irradiation environment in a nuclear reactor. The microstructure evolution before and after irradiation was systematically investigated with XRD, SEM, and TEM techniques. The results demonstrated that proton irradiation induced free volume, crystallization and H bubbles evolution. The doping of WB diminished the proton implantation dose threshold for segregation in irradiation plateaus while enhancing the growth of precipitates around the damage zone by inducing the production of M23C6 carbides and, at the same time, increasing the probability of H bubble nucleation and growth. These findings provide insights for iterative updates in Fe-based amorphous materials, informing their further development and application.

SPARC x-ray diagnostics: Technical and functional overview.

An overview is given of SPARC's three main x-ray diagnostics, with a focus on the functions they fulfill with respect to tokamak operation. The first is an in-vessel soft x-ray tomography diagnostic, aimed at providing early campaign information on plasma position, MHD activity, and impurity content. The second is an ex-vessel set of hard x-ray scintillators aimed at detecting the presence of runaway electrons, in particular during plasma startup phases. The third is a set of x-ray Bragg spectrometers, located outside of the tokamak hall, aimed at informing on the ion temperature as an indirect constraint to reduce uncertainties on the fusion power, on providing plasma rotation velocity estimates, and on observing impurity emission. Finally, more technical details are given on the beamlines at the end of which the spectrometers are located. It explains how their design allows us to ensure tritium containment and limit neutron radiation while providing a straight view into the plasma that can also be used for testing new innovative sensors.

ِِِِِِA Study of the effect of sewage water irrigation on the elemental composition of soil and plants using neutron activation analysis

ِِِِِِA Study of the effect of sewage water irrigation on the elemental composition of soil and plants using neutron activation analysis

Validation of a PGNAA sensor model for bulk material sorting

Global mineral demand is forecast to increase significantly to achieve the transition to renewable energy. Greater volumes of ore of lower grade will have to be mined to meet demand. Techniques to process large volumes of low-grade ore efficiently are being investigated to reduce the cost and impact of mining. One technique is to use sensor information to sort mined material, allowing waste to be discarded early in mineral processing. Prompt gamma neutron activation analysis (PGNAA) is a sensing technique that can provide information on the multi-elemental composition of a bulk sample which can be used for bulk material sorting. This paper presents the development of a Monte Carlo simulation model of a PGNAA sensor for bulk sorting using the Geant4 toolkit. The GEOSCAN sensor (Scantech Australia) was used as a case-study to demonstrate the application of the model. The sensor responses for a range of pure mineral samples (Fe2O3, SiO2, S, Na2CO3 and MnO2) were measured to validate the developed model. The sensitivity of the simulation results to the hadronic and electromagnetic physics models used was tested. It was determined that the PGNAA sensor model can reproduce measurements obtained from the GEOSCAN sensor. In particular, the model can provide a good reproduction of the overall spectral shape and the locations of distinct characteristic peaks. The differences between simulated and experimental results are within 30% on average. It was found that the Geant4 HP neutron model best reproduces the activation peaks observed in experimental measurements. Additionally, the PGNAA spectrum was found to be insensitive to the choice of electromagnetic model for the photon interactions. The validated sensor model provides a useful tool for investigating PGNAA sensor applications including a bulk sorting strategy for new materials, sensor calibration, improvements in signal analysis and optimised sensor design.