Neutron Activation Detectors Research Articles

The Second EWGRD Round Robin: Inter Comparison of Gamma Spectrometry Measurements on Activation Dosimeters

Following an initial Round Robin inter-comparison of gamma spectrometry measurements reported in 2014, this paper presents results from the first part of a second Round Robin inter-comparison commissioned by the European Working Group on Reactor Dosimetry in 2018. Measurements were performed by thirteen European organisations on a set of irradiated neutron activation detectors representative of those commonly used by the Reactor Dosimetry community to measure neutron fluence using gamma spectrometry methods. The radionuclides measured were110mAg,58Co,60Co,54Mn, and46Sc. The purpose of the exercise was to demonstrate the level of consistency between participating organisations in blind tests of measurements. The samples used were disks of iron, nickel, titanium, and two standard alloys of aluminium, one of 1% cobalt and the other 1% silver. They were irradiated in the MARIA reactor operated by National Centre for Nuclear Research, Poland. Participants provided their results to an independent referee who collated and compared the data. The results are presented in an anonymised form together with discussion and conclusions which may be drawn from the exercise. Good agreement was obtained with standard deviations for individual measurements between 2.0% (54Mn) and 5.6% (60Co). Overall, the results obtained from the latter Round Robin show less consistency than those from the first Round Robin, which is attributed to the participation of a wider pool of organisations.

Calculation of spectrum and neutron flux density in experimental channels of WWR-M reactor

A program for calculating the spectra and densities of neutron fluxes in the experimental channels of the research reactor WWR-M was developed. To do this, the reactor core was modeled. Irradiation of neutron activation detectors was performed. The obtained experimental rates of nuclear reactions are consistent with the calculated within 10 %.

Understanding the effects of neutron scattering for neutron-yield-isotropy measurements at the NIF.

Neutron-yield diagnostics at the NIF have been upgraded to include 48 detectors placed around the NIF target chamber to assess the DT-neutron-yield isotropy for inertial confinement fusion experiments. Real-time neutron-activation detectors are used to understand yield asymmetries due to Doppler shifts in the neutron energy attributed to hotspot motion, variations in the fuel and ablator areal densities, and other physics effects. In order to isolate target physics effects, we must understand the contribution due to neutron scattering associated with the different hardware configurations used for each experiment. We present results from several calibration experiments that demonstrate the ability to achieve our goal of 1% or better precision in determining the yield isotropy.

Design of a 16-channel fiber-optic coupled radiation detection array for radiation asymmetry studies in a 750-kJ dense plasma focus.

A 16-channel fiber optically coupled radiation detection array has been developed for studies of radiation asymmetries and emission histories from the Verus Research 750-kJ Dense Plasma Focus. Each detector in the array consists of a light-tight housing with a plastic scintillator coupled to a fiber optic that is fed into one channel of a multi-anode photomultiplier tube (PMT). The PMT and associated electronics are located in a remote electrically shielded control room. The detector head is configurable for using a Be-9 foil to take advantage of the 9Be(n,α)6He reaction as a fast neutron activation detector or with a bare scintillator alone to record the radiation emission history. Fiber optically coupling the detector head not only provides electrical isolation in the pulsed power environment but also allows the spatial footprint of the detector array to be reduced with concomitant flexibility in positioning each individual detector head. The array allows for spatially resolved neutron yield and radiation waveform measurements for fast z-pinches. The activation detector heads were calibrated for the total neutron yield against silver and indium activation counters for the total neutron yield. Fiber scintillation was found to contribute to the time-resolved detector head signals and was accounted for.

ЭКСПЕРИМЕНТЫ ПО ИССЛЕДОВАНИЮ ПРОХОЖДЕНИЯ НЕЙТРОНОВ ЧЕРЕЗ СЛОИ ГИДРИДА ЛИТИЯ-7 В УРАНОВОЙ РАЗМНОЖАЮЩЕЙ СИСТЕМЕ С ПРИМЕНЕНИЕМ МЕТОДА НЕЙТРОННО-АКТИВАЦИОННОГО АНАЛИЗА

The paper provides the precision experimental results on fission spectrum neutrons transmission through lithium hydride layers-7 with neutron activation analysis. The experiments were performed in Zababakhin All-Russia Research Institute of Technical Physics” on a critical assembly stand FKBN 2. A cylindrical multiplying system (MS) made of high enriched uranium was used as a source of fission spectrum neutrons. Activation integrals of neutron activation detectors (NADs) located at different points of the composite lithium-7 hydride reflector face-mounted on the cylindrical uranium MS were determined in the experiments. NADs based on nickel, indium, titanium, aluminum and copper were used. NADs of different types were exposed to radiation at power ~25-30 W during 1-1,5 hours. Nickel NAD was used as a monitor. Activation integrals absolute measurement error was determined by a certified technique and ranged from ~4 to ~8 % (2σ). Relative measurement error was determined by statistical accuracy, instrument peaks software processing correctness and was found not to exceed 5,4 % (2σ). Using the data obtained, spectral indices (the ratio of the normalized values of the activation integrals of various detectors types to the one of nickel detector), which provide information on the neutron spectrum were determined.

ИЗМЕРЕНИЯ ПРОХОЖДЕНИЯ НЕЙТРОНОВ С ЭНЕРГИЕЙ 14 МЭВ ЧЕРЕЗ СЛОИ ГИДРИДОВ ЛИТИЯ МЕТОДОМ ВРЕМЕНИ ПРОЛЕТА

The results of experiments on transmission of 14 MeV neutrons through lithium hydride layers of thickness up to 25 cm are presented in the article. The measurements were performed with time-of-flight method on a pulse channel of neutron generator NG-12I. The operating mode of the neutron generator is pulse-periodic. Neutrons passing through the layers of Li hydrides were registered by detector based on a 70×70 mm stilbene crystal scintillator. NIM standard modules were used as measuring equipment. The neutron yield from the generator target was estimated by neutron monitor with a fluorine plastic activation detector. Activity of radiation-exposed neutron activation detectors was measured using a gamma spectrometer with high purity germanium detector (HPGe). The averaged 14 MeV neutrons flux from the target was ∼2⋅108 n/s. The measurement results were used to obtain instrumental neutron spectra for samples of various thicknesses and to estimate coefficients of 14 MeV neutrons passing through the layers of Li hydrides. The obtained results can be used for validation of neutron-physical calculations and for improvement of neutron constants libraries.

Measuring Spectral Indices in a Critical Cylindrical Uranium System with a Lithium Hydride End Reflector

The results of precision experiments concerning neutron passage through the layers of lithium-7 hydride are reported. Activation integrals for neutron activation detectors positioned at various points of the composite lithium-7 hydride reflector mounted at the end of the uranium breeder system have been determined. Spectral indices have been calculated on the basis of the data obtained.

НЕЙТРОННО-ДОЗИМЕТРИЧЕСКОЕ СОПРОВОЖДЕНИЕ ЭКСПЕРИМЕНТОВ ПО ИЗУЧЕНИЮ СВОЙСТВ КОНСТРУКЦИОННЫХ МАТЕРИАЛОВ В РЕАКТОРНЫХ УСЛОВИЯХ

JSC "SSC RIAR" has accumulated a rich experience in the neutron-dosimetric support and neutron fields spectrometry using the activation method. Through the efforts of a specialists group, this direction continues to evolve, with advanced developments and modern high-precision equipment. The development of unified standards, computational and experimental methods of neutron-dosimetric support/monitoring to establish a one-to-one accordance between the observed changes in the properties of reactor materials and the conditions of reactor irradiation is topical. The application and distribution of neutron dosimetry using the activation method covers all the research reactors of the institute with all their various experimental conditions. The following examples of the practical neutron activation detectors usage in the following reactor materials science experiments are provided in this article.

Quantification of trace amounts of transmuted nuclides in transmutation detectors using SIMS

Neutron detectors based on neutron induced transmutation of nuclides of high-purity materials into other stable nuclides has been recently proposed to improve accuracy of long-term measurement of neutron fluence in fission nuclear reactors. The inductively coupled plasma mass spectrometry (ICP-MS) and the prompt gamma neutron activation analysis (PGNAA) have been used for the measurement of concentration of transmuted nuclides, which was then converted into neutron fluence using known neutron-reaction cross sections.In this work, the concentrations of transmuted elements were measured with a magnetic-sector secondary ion mass spectrometer (SIMS) instead of ICP-MS or PGNAA. Single crystals of Si, Ge and a polycrystalline metal foil of Ni were used as detector materials. The concentrations were measured down to sub ppb range in Si and Ge crystals. 9 prospective nuclides for transmutation detectors — 25Mg, 26Mg, 27Al, 31P, 67Zn, 69Ga, 71Ga, 59Ni, 63Ni were measured and compared to a reference neutron activation detector (NAD) method with neutron reaction cross section values taken from two different databases. With exception of 26Mg, the values determined using SIMS were within 20 % compared to those obtained using NADs with a cross section value from one or both databases. In case of 26Mg the difference between the results of SIMS and NAD methods was higher than 45 % for data from both databases. This discrepancy for 26Mg measurement was assigned tentatively to uncertainties of the neutron reaction cross section values. Nevertheless, the experiments proved distinct advantages of SIMS over ICP-MS in sample preparation and throughput, and in consuming micro-volumes of material and identified at least one well performing transmutation detector for thermal neutrons (Ge - single crystal) with agreement within 3 % between SIMS-TMD and NADs.

ИЗМЕРЕНИЯ ПРОХОЖДЕНИЯ НЕЙТРОНОВ С ЭНЕРГИЕЙ 14 МэВ ЧЕРЕЗ СЛОИ ГИДРИДОВ ЛИТИЯ МЕТОДОМ НЕЙТРОННО-АКТИВАЦИОННОГО АНАЛИЗА

The paper presents the results of experiments in 14-MeV neutron transmission through lithium hydride layers 25 cm thick. Measurements were performed by neutron-activation analysis using NG-12I neutron generator. The neutron generator operated in static mode. In order to determine 14-MeV neutron transmission through sample layers and to evaluate neutron spectrum changes we have used neutron-activation detectors based on the following activation reactions: 19F(n,2n)18F, 27Al(n,α)24Na, 64Zn(n,p)64Cu, 115In(n,n')115mIn, 65Cu(n,2n)64Cu, 63Cu(n,γ)64Cu. The activity of irradiated neutron-activation detectors was measured by the gamma-spectrometer with a semiconductor HPGe-detector. Neutron yield from the generator target was determined by monitor readings with fluoroplastic neutron-activation detector. The average 14-MeV neutron flow from the target came up to ~1010 n/s. The experimental geometry and measurement results are described. Based on measurement data we determined the number of reactions for each neutron-activation detector, as well as coefficients of 14-MeV neutron transmission through lithium hydride layers of different thickness. The obtained results can be used to validate neutron-physical calculations and to update neutron constant libraries.

A.161 - Neutron activation detectors for clinical dosimetry in external beam radiation treatments

A.161 - Neutron activation detectors for clinical dosimetry in external beam radiation treatments

The EWGRD Round Robin Measurement Exercise

This paper presents the results of a round robin exercise carried out to compare specific activity measurements performed by eight European organisations on a set of ten neutron activation detectors containing the radio-nuclides 110m Ag, 60 Co, 54 Mn, 46 Sc and 94 Nb. The purpose of the exercise was to demonstrate the level of consistency between the participating organisations in blind tests of measurements relevant to reactor metrology. The samples used were selected from a stock of pre-existing irradiated material held at SCK•CEN. Taking turns over a period of approximately 9 months, the participating organisations received the samples, measured them and provided their results to an independent referee who collated and compared the data. The inter-comparison has demonstrated good agreement between the participants with standard deviations for each dosimeter varying between 1.6% and 3.1%. The paper provides results of the EWGRD Round Robin in an anonymised form together with discussion and conclusions which may be drawn from the exercise.

Experimental investigation of neutronic characteristics of the IR-8 reactor to confirm the results of calculations by MCU-PTR code

A comparison of measured and calculated neutronic characteristics (fast neutron flux and fission rate of 235U) in the core and reflector of the IR-8 reactor is presented. The irradiation devices equipped with neutron activation detectors were prepared. The determination of fast neutron flux was performed using the 54Fe (n, p) and 58Ni (n, p) reactions. The 235U fission rate was measured using uranium dioxide with 10% enrichment in 235U. The determination of specific activities of detectors was carried out by measuring the intensity of characteristic gamma peaks using the ORTEC gamma spectrometer. Neutron fields in the core and reflector of the IR-8 reactor were calculated using the MCU-PTR code.

Locally manufactured films for neutron flux measurement in the MNSR type reactor

Metal films deposited on Teflon are used in the Miniature Neutron Source Reactor (MNSR) for the first time to study their usability as neutron activation detectors for the thermal neutron flux measurements in the reactor. For this purpose Titanium, Aluminum, Vanadium, and Silver films deposited on Teflon have been locally prepared at room temperature using two methods: the vacuum arc deposition and DC Magnetron sputtering techniques. The thermal neutron flux in the MNSR inner irradiation site was measured using the prepared metal films. The results at the 95% level of confidence of the neutron flux using the metal films deposited on Teflon by the vacuum arc deposition for Titanium, Aluminum, and Vanadium were: (9.9±0.3)×1011, (1.4±0.3)×1012, (1.2±0.2)×cm−2s−1, respectively. The result at the same level of confidence of the neutron flux using the metal films deposited on Teflon by the DC Magnetron sputtering for Silver was: (1.5±0.2)×1011cm−2s−1. Good agreements are noticed between our obtained mean value (9.3±0.9)×1011cm−2s−1 and the previous published results.

Predicting the sensitivity of the beryllium/scintillator layer neutron detector using Monte Carlo and experimental response functions.

A methodology for obtaining empirical curves relating absolute measured scintillation light output to beta energy deposited is presented. Output signals were measured from thin plastic scintillator using NIST traceable beta and gamma sources and MCNP5 was used to model the energy deposition from each source. Combining the experimental and calculated results gives the desired empirical relationships. To validate, the sensitivity of a beryllium/scintillator-layer neutron activation detector was predicted and then exposed to a known neutron fluence from a Deuterium-Deuterium fusion plasma (DD). The predicted and the measured sensitivity were in statistical agreement.

Radiometric Complex with Gamma Spectrometers for Measuring the Characteristics of Neutron Fields at Nuclear-Physics Installations

A method for measuring the induced activity in neutron activation detectors using gamma spectrometers with scintillator and semiconductor detectors is introduced. The sensitivity of the gamma spectrometers is measured as a function of γ-ray energy over a range of 0.06-3 MeV.

Performance of an improved thermal neutron activation detector for buried bulk explosives

First generation thermal neutron activation (TNA) sensors, employing an isotopic source and NaI(Tl) gamma ray detectors, were deployed by Canadian Forces in 2002 as confirmation sensors on multi-sensor landmine detection systems. The second generation TNA detector is being developed with a number of improvements aimed at increasing sensitivity and facilitating ease of operation. Among these are an electronic neutron generator to increase sensitivity for deeper and horizontally displaced explosives; LaBr3(Ce) scintillators, to improve time response and energy resolution; improved thermal and electronic stability; improved sensor head geometry to minimize spatial response nonuniformity; and more robust data processing. The sensor is described, with emphasis on the improvements. Experiments to characterize the performance of the second generation TNA in detecting buried landmines and improvised explosive devices (IEDs) hidden in culverts are described. Performance results, including comparisons between the performance of the first and second generation systems are presented.

Calculations to support JET neutron yield calibration: Neutron scattering in source holder

After the coated CFC wall to ITER-Like Wall (Beryllium/Tungsten/Carbon) transition in 2010–11, confirmation of the neutron yield calibration will be ensured by direct measurements using a calibrated 252Cf neutron source deployed by the in-vessel remote handling boom and Mascot manipulator inside the JET vacuum vessel.The paper describes preliminary calculations and the results of numerical study of the effect of source holder on neutron detector response. The source baton was designed in such a way, that it does not significantly affect the neutron spectrum, angular neutron flux distribution or activation detector response. All effects are approximately equal to or less than 1%. The largest disturbance to the neutron flux angular distribution and to the neutron spectrum arises from the source capsule. Hence one should obtain as much information as possible about the capsule and the 252Cf source material in order to avoid additional systematic errors.

The Fast Neutron Fluence and the Activation Detector Activity Calculations Using the Effective Source Method and the Adjoint Function

Abstract This paper describes the application of effective source in forward calculations and the adjoint method to the solution of fast neutron fluence and activation detector activities in the reactor pressure vessel (RPV) and RPV cavity of a VVER-440 reactor. Its objective is the demonstration of both methods on a practical task. The effective source method applies the Boltzmann transport operator to time integrated source data in order to obtain neutron fluence and detector activities. By weighting the source data by time dependent decay of the detector activity, the result of the calculation is the detector activity. Alternatively, if the weighting is uniform with respect to time, the result is the fluence. The approach works because of the inherent linearity of radiation transport in non-multiplying time-invariant media. Integrated in this way, the source data are referred to as the effective source. The effective source in the forward calculations method thereby enables the analyst to replace numerous intensive transport calculations with a single transport calculation in which the time dependence and magnitude of the source are correctly represented. In this work, the effective source method has been expanded slightly in the following way: neutron source data were performed with fewgroup method calculation using the active core calculation code moby-dick. The follow-up neutron transport calculation was performed using the neutron transport code tort to perform multigroup calculations. For comparison, an alternative method of calculation has been used based upon adjoint functions of the Boltzmann transport equation. Calculation of the three-dimensional (3-D) adjoint function for each required computational outcome has been obtained using the deterministic code tort and the cross section library BGL440. Adjoint functions appropriate to the required fast neutron flux density and neutron reaction rates have been calculated for several significant points within the RPV and RPV cavity of the VVER-440 reactor. Both of calculation methods are briefly described and the results are compared.

Beryllium neutron activation detector for pulsed DD fusion sources

A compact fast neutron detector based on beryllium activation has been developed to perform accurate neutron fluence measurements on pulsed DD fusion sources. It is especially well suited to moderate repetition-rate (<0.2 Hz) devices, such as the plasma focus or Z-pinch. The detector comprises a beryllium metal sheet sandwiched between two large-area xenon-filled proportional counters. A methodology for calculating the absolute response function of the detector using a “first principles” approach is described. This calibration methodology is based on the 9 Be(n,α) 6 He cross-section, energy calibration of the proportional counters, and numerical simulations of neutron interactions and beta-particle paths using MCNP5. The response function R ( E n ) is determined over the neutron energy range 2–4 MeV. The count rate capability of the detector has been studied and the corrections required for high neutron fluence measurements are discussed. For pulsed DD neutron fluencies >3×10 4 cm −2 , the statistical uncertainty in the fluence measurement is better than 1%. A small plasma focus device has been employed as a pulsed neutron source to test two of these new detectors, and their responses are found to be practically identical. Also the level of interfering activation is found to be sufficiently low as to be negligible.