Foil Activation Research Articles

Design and validation of a single cylindrical neutron spectrometer using a gold activation foil

A single cylindrical neutron spectrometer (SCNS) with a gold activation foil was designed for mono-directional high-flux neutron spectrum measurement applications. A total of seven gold activation foils with a diameter of 11 mm and a thickness of 0.2 mm were placed in the moderator as detectors with a depth variation of 15 mm. The moderator was cylindrical with a diameter of 8 cm and a length of 16 cm and made of polyethylene. Outside the moderator is a neutron shield made of 3 mm boron-rubber and 0.5-mm cadmium and covered with an aluminum casing of 5 mm thickness. The response value of each detector was determined by simulation using the Monte Carlo MCNPX 2.7 program. The simulated response detectors was then partially validated by comparing the simulation and the measurement result with an AmBe neutron source, where the difference is around 3%. The obtained results showed that SCNS with a gold activation foil could be used to measure the neutron spectrum in a high-flux neutron field.

Simulation of the DD, TT and DT JET Neutron Streaming Experiments with TRIPOLI-4® Monte-Carlo code

The DD, TT and DT neutron streaming experiments (NEXP-STRE) in the fusion facility JET are here simulated with TRIPOLI-4® Monte-Carlo code. Specifically, we present the results from the simulation of the DD (2019) and the later TT and DT (2021) campaigns. The last version of TRIPOLI-4® code, developed by CEA, is provided with an innovative variance reduction technique based on the adaptive multi-level splitting (AMS) methodology for particle transport which ensures reliable results in very detailed, complex and vast geometry. Furthermore, a specific tool, developed by CEA, is used to automatically convert MCNP geometry into TRIPOLI-4® geometry format, which allows to perform fast and reliable comparisons between the results obtained by the mean of both calculation tools. The JET 360° MCNP model is used as a basis for creating the TRIPOLI-4® geometry. The neutron flux and doses in thermoluminescent detectors (TLDs) and reaction rates in activation foil (AF) detectors are computed for various positions in the experimental hall. A comparison of our results with the experimental values (2019 DD - 2021 TT campaigns) and the MCNP computed values (all campaigns) is presented. The AMS results crucial for obtaining statistically reliable quantities with TRIPOLI-4® code.

Experimental evaluation of transition rate of sapphire crystal for thermal and fast neutrons using MNSR vertical neutron beam line

Using a perfect single crystal as a neutron filter allows us to have a thermal neutron beam with almost no background of fast neutrons. Single crystals of Al2O3 (sapphire) have proven to be effective filters for fast neutrons and are incorporated into neutron instruments. The present work would experimentally investigate c-axis neutron transmission rate by using different crystal thicknesses. In fact, the optimal thickness for sapphire filter is the one that maximizes the transmission of low energy neutrons and minimizes the transmission of fast neutrons, if there is no significant decrease in thermal neutron flux. In addition, neutron-filtering power of a-axis and c-axis sapphire crystals were compared with each other using different tests on a 2.5 cm slab of the sapphire crystals. The experimental tests were carried out by means of the available neutron flux top of the vertical neutron beam line of the Isfahan Miniature Neutron Source Reactor (MNSR) in two methods of foil activation and flux monitoring. In addition, the thermal and fast neutron dose rate reduction was discussed by using different thicknesses of the c-axis crystal.

Mutual shielding of neutron activation foils

For decommissioning purposes, stacks of foils are used in order to determine the neutron fluence rates in reactor buildings. Due to the relatively great thickness of these foils, self-shielding, as well as mutual shielding become significant for these stacks. In this paper, the results of Monte Carlo calculations are presented for the self-shielding of single foils, and also for the mutual shielding of foils in a stack. Because the irradiation geometry in a reactor building at a large distance from the reactor core is probably not isotropic, especially not so for neutrons of higher energy, the shielding effects were calculated for both isotropic and mono-directional irradiation geometries. The results obtained show a number of general trends which can be understood with a few rules of thumb.

Neutron source strength verification via reaction rate measurement and Monte Carlo simulation

Determining the neutron source strength is essential to ensure optimal use of the source and to implement adequate radiation protection measures. In this study, the source strength of two (α,n) neutron sources – a newly acquired 185 GBq 241AmBe and a legacy source 74 GBq 239PuBe – were verified. Source strength verification was performed by measuring the induced activities in irradiated 115In foils and comparing the results to MCNP calculated reaction rates. The results for 241AmBe agree well with a maximum and average deviations of 8.1% and 4.1%, respectively. While for 239PuBe the maximum deviation is 11.7% and the average deviation is 6.3%. Results of this study provide a reliable reference for the application of the sources, in particular, for the 241AmBe which is used as the external neutron source of the PRR1-Subcritical Assembly for Training, Education, and Research (PRR1-SATER), and for the design of a neutron irradiator that will repurpose the 239PuBe.

Measurements of proton beam flux and energy of APEP using foil activation technique

The activation method of metallic foils is an important technique to measure the flux and energy of proton beams. In this paper, the method was used to measure the CSNS APEP proton flux at seven nominal proton energies ranging from 10 MeV to 70 MeV for beam spot sizes of the 20 mm × 20 mm and 50 mm × 50 mm. The reactions of natTi(p, x)48V, natNi(p, x)57Ni, natCu(p, x)58Co, and 27Al(p, x)24Na were employed to measure the proton beam flux with a range of 107-109 p/cm2/s. Furthermore, we also proposed a method using the activity ratio with a stacked-foil target to determine the energy spread of a Gaussian-like distribution for different nominal proton energies. The optimal combinations of Al, Cu, Ti, Ni, Mo, Fe, Nb, and In foils were adopted for the proton energies. The measured energy spreads for degraded beams of 30 MeV–70 MeV were found to be smaller than 10.00%.

A Monte Carlo model of the Dingo thermal neutron imaging beamline

In this study, we present a validated Geant4 Monte Carlo simulation model of the Dingo thermal neutron imaging beamline at the Australian Centre for Neutron Scattering. The model, constructed using CAD drawings of the entire beam transport path and shielding structures, is designed to precisely predict the in-beam neutron field at the position at the sample irradiation stage. The model’s performance was assessed by comparing simulation results to various experimental measurements, including planar thermal neutron distribution obtained in-beam using gold foil activation and ^{10}B_{4}C-coated microdosimeters and the out-of-beam neutron spectra measured with Bonner spheres. The simulation results demonstrated that the predicted neutron fluence at the field’s centre is within 8.1% and 2.1% of the gold foil and ^{10}B_{4}C-coated microdosimeter measurements, respectively. The logarithms of the ratios of average simulated to experimental fluences in the thermal (E_{th}< 0.414 eV), epithermal (0.414 eV < E_{epi}< 11.7 keV) and fast (E_{fast}> 11.7 keV) spectral regions were approximately − 0.03 to + 0.1, − 0.2 to + 0.15, and − 0.4 to + 0.2, respectively. Furthermore, the predicted thermal, epithermal and fast neutron components in-beam at the sample stage position constituted approximately 18%, 64% and 18% of the total neutron fluence.

Thermal neutron capture cross-section and resonance integral measurements of 186W(n,γ)187W reaction using the thermal column neutron source at the Dalat research reactor

The thermal neutron radiative capture cross-section (σ0) and the resonance integral (I0) of the reaction 186W(n,γ)187W were measured by the thin foil activation method using the well thermalized neutron source at the graphite thermal column facility of the Dalat research reactor. The activities of the irradiated target and monitor samples were measured using a high resolution HPGe gamma-ray spectrometer. The recent updated values of gamma-rays intensities and decay half-life of the 187W nucleus were applied in this experiment to substantially improve the cross-section results compared to the previous measurements. The thermal neutron capture cross- section and the resonance integral of the 186W(n,γ)187W reaction were determined relative to the standard reaction 197Au(n,γ)198Au. The values obtained are 31.76 ± 0.63 and 429.2 ± 25.6 barns for σ0 and I0 respectively. Our results and previous measured values are compared, and discrepancies are discussed.

Measurement of alpha-induced reaction cross-sections for [formula omitted]Zn with detailed covariance analysis

The production cross-section of 68Ge, 69Ge, 65Zn and 67Ga radioisotopes from alpha-induced nuclear reaction with natZn have been measured using the stacked foil activation technique followed by the off-line γ-ray spectroscopy in the incident alpha energy range 14-37 MeV. The obtained nuclear reaction cross-sections are compared with previous experimental data available in the EXFOR data library, evaluated nuclear data from TENDL-2019 and theoretical results, calculated using TALYS nuclear reaction code. We have also performed the detailed uncertainty analysis for these nuclear reactions and their respective correlation metrics are presented. Since α-induced reactions are important in nuclear medicine and developing the nuclear reaction codes so needful corrections related to the coincidence summing factor and the geometric factor have been considered during the data analysis in the present study.

Experimental Analyses of Capture Reaction Rates for Epithermal and Resonance Neutrons in Source-Driven Subcritical Cores

Neutron irradiation experiments are carried out in source-driven subcritical cores with high-energy neutrons generated by spallation reactions between a 100-MeV proton beam and a lead-bismuth target at the Kyoto University Critical Assembly. The main objective of the experiments is to investigate the effect of epithermal and resonance neutrons on the accuracy of capture reaction rates with respect to a subcriticality variation. Activation foils of copper, indium, tantalum, and tungsten are employed to obtain capture reaction rates for epithermal and resonance neutrons by applying the cadmium difference method. Also, the applicability of the foils for the measurement of the reaction rates for epithermal and resonance neutrons is substantiated in the critical irradiation experiments performed prior to the subcritical experiments. The subcritical experiments are conducted with three different subcriticalities by changing the control rod insertion pattern. The measured reaction rates are compared with the calculated values obtained by the Monte Carlo code MVP with JENDL-4.0, and the ratio of the calculation and experiment values of the reaction rates shows equivalent values within the 1σ errors regardless of a difference in the subcriticality. The compared results indicate that the numerical analyses have a consistent accuracy of reaction rates in epithermal and resonance energy regions for a subcriticality variation in source-driven subcritical cores.

First bromine doped cryogenic implosion at the National Ignition Facility

We report on the first experiment dedicated to the study of nuclear reactions on dopants in a cryogenic capsule at the National Ignition Facility (NIF). This was accomplished using bromine doping in the inner layers of the CH ablator of a capsule identical to that used in the NIF shot N140520. The capsule was doped with 3 × 1016 bromine atoms. The doped capsule shot, N170730, resulted in a DT yield that was 2.6 times lower than the undoped equivalent. The Radiochemical Analysis of Gaseous Samples (RAGS) system was used to collect and detect 79Kr atoms resulting from energetic deuteron and proton ion reactions on 79Br. RAGS was also used to detect 13N produced dominantly by knock-on deuteron reactions on the 12C in the ablator. High-energy reaction-in-flight neutrons were detected via the 209Bi(n,4n)206Bi reaction, using bismuth activation foils located 50 cm outside of the target capsule. The robustness of the RAGS signals suggests that the use of nuclear reactions on dopants as diagnostics is quite feasible.

Neutron Flux Determination for the NEAR Station at the CERN n_TOF Facility Using the SAND II Unfolding Code

The NEAR station, a new experimental area of the n_TOF facility, was established after the second long shut down of CERN in 2020. This new area was created in order to exploit in situ the high instantaneous neutron flux (originating from proton spallation into a lead target in bunches with momenta reaching up to 20 GeV/c). The reason for the creation of the NEAR station is to utilize the neutrons for experiments related to radiation damage on materials and nuclear astrophysics and the corresponding neutron flux was characterized via extensive neutron multiple foil activation measurements. The irradiated foils were subsequently measured using high purity germanium detectors (HPGe) to determine their induced activities. Finally, the widely used SAND II unfolding code was implemented for the characterization of the neutron flux, using the evaluated cross sections from the IRDFF library, along with the experimentally derived activities. The preliminary results concerning the neutron flux determined in the present work are compared with the corresponding FLUKA simulated ones.

Towards the use of SDD as an absolute detector for high-energy neutron measurements

As of today, the standard method employed in tokamaks for the absolute measurement of the neutron flux (thus of the nuclear fusion power) is based on activation foils, being the most robust and unbiased technique for the absolute determination of neutron fluence. However, this technique is not able to provide real-time data useful for the control of future fusion plants like DEMO. In this paper, we present some preliminary results about the R&D activity aimed at developing the Single-crystal Diamond Detectors (SDD) used for fast neutron measurements into an absolute neutron flux monitor. Tests have been conducted at the new NILE neutron source of the Rutherford-Appleton Laboratory, a facility with compact neutron generators with a maximum yield of 109 n/s and 1010 n/s for 2.5 MeV and 14 MeV neutrons, respectively. A series of neutron spectra and flux measurements have been taken with different SDD and associated DAQ. Comparisons with standard activation foils (and namely Fe, Zr, Al and Nb foils for 14 MeV neutrons and In for 2.5 MeV neutrons) and with other reference detectors are presented and discussed. Also discussed is the stability of the SDD over time when employed at high neutron rates in realistic neutron environment, and the effects of neutron irradiation on both the counting rate and detector resolution.

FOIL ACTIVATION TECHNIQUE-A TOOL FOR THE EVALUATION OF PHOTO-NEUTRON DOSE IN RADIOTHERAPY.

Treatment of cancer is carried out using photon beams from high-energy medical linear accelerators. Photo-neutrons are also produced as an unwanted by product in the process of dose delivery to the cancer patients during their radiation treatments. In the present study, photo-neutron dose equivalents (both thermal and fast components) per unit delivered gamma-photon dose were measured at different depths, as function of distances from iso-centre in patient plane, field sizes, wedge angles and at LINAC head for a 15-MV medical linear accelerator model Elekta Precise using multi-foil activation technique. The neutron dose equivalents determined for the above-mentioned parameters were found to be lower (<0.05%) in comparison with the therapeutic photon dose delivered and within the prescribed limits recommended by the national regulatory authority.

Characterization of thermal neutron distribution of an Am–Be neutron source setup by CdZnTe detector

Owing to the developments in prompt gamma neutron activation analysis (PGNAA) and prompt gamma ray activation imaging (PGAI), it is necessary to develop an online thermal neutron distribution measurement method. The CdZnTe detector is considered as an alternative thermal neutron detector because of its high thermal neutron capture cross section. In this study, the thermal neutron field of an 241Am–Be neutron source was determined by CdZnTe detector. The intrinsic neutron detection efficiency of CdZnTe detector was calculated by using indium foil activation, and the value was 3.65%. Then the characteristics of neutron source were conducted with the calibrated CdZnTe detector. The thermal neutron fluxes in front of beam port were measured at several distances ranging from 0 cm up to 28 cm. Thermal neutron field at distances of 1 cm and 5 cm were also measured. The experimental data were then compared with Monte Carlo simulation. The results demonstrated that the simulated data agree well with experimental measurements.

Probing reaction mechanisms through residual cross section measurement for 7Li+Zn system

Though the picture remains incoherent to draw a concrete conclusion, especially with weakly bound particles, prolific efforts have been devoted to unfolding the reaction mechanisms at energies below 10 MeV/u. The dearth of extensive data and ambiguous findings in the light to medium mass region galvanizes the exploration of the entire mass domain. In the course of discerning the underlying reaction mechanisms in 7Li fusion with natural Zn, the excitation function (EF) analysis has been reported within the energy range of 21–43 MeV. Residual cross sections have been measured employing the stacked foil activation technique followed by offline γ-spectroscopy. Subsequently, the measured EFs have been compared with theoretical predictions from statistical model codes PACE4 and EMPIRE3.2.2. EF analysis affirms the dominance of the complete fusion (CF) mechanism as revealed through the gross satisfaction of measured data from the n-emitting channels by both the model codes. The observed enhancement in subsequent channels relative to theory has been implicitly interpreted as the role of other mechanisms like breakup fusion and/or direct cluster transfer in populating residues owing to the low breakup threshold of weakly bound 7Li. A critical piece of experimental evidence concerning the discrepancy in the branching intensity of 93.31 keV γ-peak from the decay of 67Ga residue reported in various nuclear databases has been witnessed. Thus, the revised estimated intensity value has been reported, along with a discussion on the optimized production of Ga isotopes for medicinal applications.

105Rh yield from the proton induced fission of uranium

The yield of R105u and R105h produced through proton induced fission of U238 are measured using stacked foil activation technique. The results are compared with theoretical calculations using TALYS 1.95. As the R105h shows a higher cross-section for neutron absorption in the thermal region, production of this isotope will act as a reactor poison. Hence the reactivity equivalent of R105h is also calculated and is found to be −0.84%.

Production of 11C, 13N and 15O in proton-induced nuclear reactions up to 200 MeV

Production cross sections of light positron emitters 11C (t1/2=20.36 min), 13N (t1/2=9.97 min) and 15O (t1/2=122 s) have been measured for proton-induced reactions on the main relevant constituents of the human body (C, N and O) from threshold up to 200 MeV of proton incident energy. This has been accomplished by means of two complementary experiments at the clinical proton treatment centers WPE (Essen) and HIT (Heidelberg), both in Germany. In the first case, the multi-foil activation technique was combined with PET imaging, a methodology previously developed at CNA (Seville, Spain). In the second experiment, aimed at energies below 50 MeV, the activation of single foils was measured with conventional LaBr3 detectors. In both cases the IAEA monitor reaction natCu(p,x)63Zn data was used for validation purposes.The most complete and accurate set of excitation functions for the production of light-positron emitters 11C, 13N and 15O in proton induced reactions on human tissue up to 200 MeV is presented. These data are needed for nuclear reaction model development and are extremely important to improve the accuracy of the PET range verification in proton radiotherapy. Seven excitation functions have been measured: 12C(p,x)11C, 14N(p,x)11C, 14N(p,x)13N, 14N(p,γ)15O, 16O(p,x)11C, 16O(p,x)13N and 16O(p,x)15O up to 200 MeV of proton incident energy.

Measurement of delayed neutrons in a thermal nuclear reactor by means of a long run pile noise experiment in sub-critical state

The transfer function of a zero-power thermal reactor was successfully measured thanks to neutron noise techniques from 1 mHz to 160 Hz. During a month-long experimental campaign, the fluctuations of the neutron population in critical and subcritical configurations of the core were acquired using excore fission chambers and analysed through Cross-Power Spectral Density (CPSD) methodology. Firstly, the reactor’s kinetic parameters, i.e. prompt decay constant, effective delayed neutron fraction and generation time, were obtained at critical state. It required calibrating the reactor’s power, which was done by metal foil activation and measurement of the 235U fission rate. Secondly, these parameters were used to estimate the groups’ abundances of delayed neutrons from the CPSD measured in a sub-critical state. Fitting data with a point kinetic model was done with Bayesian inference - CONRAD and Stan programs were used. A very good agreement was found between experimental abundances and the ones computed with TRIPOLI-4 Monte-Carlo transport code and JEFF3.1.1 nuclear data library. Uncertainties on prior abundances between 6 % to 101 %, held mainly by nuclear data, were lowered down to 4 % to 54 %, depending on the delayed group.

Validation of Neutron Flux and Activation in the Nagra AMAC model of the Mühleberg Power Plant

The decommissioning of the Mühleberg Nuclear Power Plant (KKM) is supported by Nagra (National Cooperative for the Disposal of Radioactive Waste) through computational analysis of neutron activation in the reactor components and structures using the Nagra A dvanced M ethodology for A ctivation C haracterization (AMAC). To validate these calculations, measurements of neutron flux and neutron-induced activity were carried out in the form of two foil activation campaigns (72 samples) and three concrete sampling campaigns (118 samples). Generally, the ratio of calculated to measured activities is in the order of 2 in the proximity of the reactor pressure vessel and 3 in the upper sections of the drywell. As expected, the flux in the lower (spherical) part of the drywell is significantly overestimated due to the simplified modelling of this space. The overestimation, which ratios of up to 10, can be corrected by scaling the calculated radionuclide activity distributions in the vicinity of the validation locations.