Thermal Neutron Fluence Rate Research Articles

A compact moderator for thermal neutron reference radiation field with an 241Am-Be source

To calibrate thermal neutron detectors, State Key Laboratory of NBC Protection for Civilian plans to establish thermal neutron reference radiation field (TNRRF) based on the existing 7.4 × 1011 Bq 241Am-Be source. An 82 cm (L) × 50 cm (W) × 50 cm (H) moderator was designed through Monte Carlo simulation to thermalize fast neutrons. Each component was determined to optimize thermal neutron fluence rate φ, proportion of thermal neutrons ρ, and uniformity factor μ of the radiation field. A 3He proportional counter was used to evaluate the fabricated moderator. Results show that surrounding reflector is the main origin of thermal neutrons. D2O tank can promote ρ and its length can be utilized to adjust φ. Homogenizing lens will improve μ and reduce φ. Experimental φ, cadmium ratio η, and μ at the reference position are 1808.82 ± 85.96 cm−2 s−1, 315.80, and 5.29 %, respectively, which are consistent with simulated results. Next, TNRRF will be established after long-term stability testing.

Activation level of the concrete building and pressure vessel in JAEA-Tokai tandem accelerator

ABSTRACT The activation level of the JAEA-Tokai tandem accelerator facility was investigated experimentally in advance of the future decommissioning. JAEA-Tokai tandem accelerator facility has a higher terminal voltage of 18 MV and a larger total floor area, Compared to other electrostatic accelerators with terminal voltage of 1 MV to 6 MV. Therefore, determination for ‘where,’ ‘what,’ and ‘how many’ nuclides are produced in the facility is crucial. Thermal neutrons generated by beam losses associated with accelerator operations contribute significantly to the activation of equipment and facilities. The accumulated activities of60Co and152Eu; the most considerable radionuclides at the decommissioning, can be deduced by the thermal neutron fluence rate during the accelerator operation. In this study, thermal neutron fluence measurement on the surface of the pressure vessel and concrete building was conducted with conventional methods using dosimeters and metal foil detectors as well as a new method using a portable γ-ray detector. The thermal neutron fluence in the facility during the accelerator operation ranges from 101 to 104 n/cm2/s. The sum of deduced activities of60Co and152Eu in 50 years is much lower than the clearance level of 0.1 Bq/g in all areas except in the irradiation room.

Thermal neutron reference radiation facility with high thermalization and large uniformity area

A new thermal neutron facility based on 12 241Am–Be neutron sources has been established at the National Institute of Metrology, China. It has two independent irradiation fields, the inner field and outer field, constructed with high-purity graphite and heavy water as moderators, respectively. Three innovative designs, including the reflection cavity, reflection layer, and homogenizing lens, were introduced to improve the performance of the facility. The reflection layer can increase the thermal neutron fluence rate by about 3–4 times, the reflection cavity also enhances thermal fluence by 3–4 times by taking advantage of multiple scattering of neutrons, and the homogenizing lens mounted on the neutron exit surface of the moderator improve the distribution of thermal neutron. The characteristics of the two fields were investigated by Monte Carlo simulations and experimental measurements. The results show that the new facility has high thermalization and large uniformity area with these innovative designs. For the inner field, the thermal neutron fluence rate at the reference position is (21 433.3 ± 407.2) cm−2 s−1, the cadmium ratio is 20.6, the non-uniformity is 0.1% over the 40 cm × 40 cm region. For the outer field, the thermal fluence rate at the reference position is (2046.0 ± 49.1) cm−2 s−1, the cadmium ratio is 1433, and the non-uniformity is 0.7% in the 70 cm × 70 cm region. The facility is expected to be suitable for the large-size detector, and its excellent performance is attractive for various thermal neutron experiments.

Estimation with Monte Carlo of thermal neutrons in the FANT, using 252Cf and 241Am/9Be neutron source

The thermal neutron irradiation device (FANT), developed at the Neutron Measurements Laboratory of the Energy Engineering Department at Universidad Politécnica de Madrid, is a high-density polyethylene regular parallelepiped, with a rather uniform neutron fluence inside its irradiation chamber. It uses a Am95241/Be49 neutron source aiming to provide thermal neutron fluence rates. Neutron spectra and neutron fluences were estimated with Monte Carlo methods in the FANT irradiation chamber when a Cf98252 neutron source is used and were compared with the results obtained with the Am95241/Be49 source. Regardless of the neutron source, the largest contribution is due to thermal neutrons, producing also epithermal and fast neutrons. Per neutron emitted by the source, the use of Cf98252 produces a larger amount of neutrons.

Comparison of FANT results using the ENDF/B-VII.1, JEFF-3.3 and TENDL2017 nuclear data libraries

FANT (Fuente Ampliada de Neutrones Térmicos; in Spanish) is a thermal neutron irradiation facility with an extended and very uniform irradiation area, that has been developed by the Neutron Measurements Laboratory of the Energy Engineering Department at Universidad Politecnica de Madrid (LMN-UPM). This device is a parallelepiped box made of high-density polyethylene (HDPE), moderator material, that uses an A95241m/B49e neutron source of 111 GBq nominal activity for irradiating materials. The facility design was previously optimized, and the neutron spectra were estimated by extensive calculations with the MCNP6.1 code and carrying out experimental measurements (Bedogni et al., 2017). The facility takes advantage of the scattering reactions of neutrons with the HDPE surfaces of the chamber, where the moderation process is effective, achieving relevant thermal neutron fluence rates. The main goal of this work has been to simulate and analyse the FANT system by Monte Carlo methods using the MCNP6.1 code, employing 3 different nuclear data libraries: ENDF/B-VII.1, JEFF-3.3 and TENDL 2017. The transport of thermal neutrons in HDPE, E < 1eV, has been calculated in all the cases taking into account the thermal S (α,β) treatment. The results achieved in this work have been compared with those previously obtained in the former development of FANT, using the MCNP6.1 code with the ENDF/B-VII.1 nuclear data, and experimental measurements. These results have shown that the JEFF-3.3 nuclear data library is the nuclear data library that provides of the best matching between the MCNP computational results, and the experimental data collected at FANT. Hence, the JEFF-3.3 nuclear data library seems to be the most correct library to design and benchmark thermal neutron activation devices.

Experimental characterization of FANT, a new thermal neutron source

FANT is the acronym of Enhanced Thermal Neutron Source (Fuente Ampliada de Neutrones Térmicos, in Spanish). This is a parallelepiped box of high-density polyethylene moderator and an isotopic neutron source. The moderator has a cylindrical irradiation chamber where a rather uniform thermal neutron flux is obtained. The FANT design was previously optimized and the neutron spectra were estimated by Monte Carlo calculations with the MCNP6.1 code. To check the characteristics of the FANT thermal neutron field, measurements have been performed at the reference point inside the irradiation chamber with a Bonner sphere spectrometer holding a small 6LiI(Eu) thermal neutron detector. To unfold the neutron spectrum BUNKIUT with UTA4 response matrix and NSDann Ver 4.0 codes were used. Some issues have been found and recommendations are made about the use of large BSS inside narrow spaces, and about the capacity of NSDann code to unfold these kind of spectra. However, the results confirm that the moderation process in FANT is very effective and allows obtaining useful thermal neutron fluence rates.

The e_LiBANS facility: A new compact thermal neutron source based on a medical electron LINAC

A new photonuclear thermal neutron facility has been developed at the Physics Department of University of Torino. The facility is based on a medical electron LINAC coupled to a compact converter and moderator assembly. A homogeneous thermal neutron radiation field of the order of 106 cm−2s−1 is achievable in the enclosed irradiation cavity with low gamma and fast neutron contaminations. Its intensity can be tuned varying the LINAC current. These characteristics make the source appropriate for several applications such as detectors development, material studies and BNCT preclinical research. This work describes the project and the experimental characterization of the facility. This includes the measurement of the thermal neutron fluence rate, the determination of the neutron energy spectrum and of the thermal neutron field uniformity and the evaluation of the gamma contamination.

Reference thermal neutron field at KRISS for calibration of neutron detectors

A reference thermal neutron field has been established at the Korea Research Institute of Standards and Science (KRISS) by using a 241Am-Be neutron source and a high-purity graphite pile constructed by stacking graphite blocks. The properties of the graphite blocks such as impurities, density, and dimensions were studied thoroughly to understand the characteristics of the generated field. The energy spectrum and thermal neutron fractions were simulated with the Monte Carlo N-Particle eXtended code using measured physical parameters. The neutron effective temperature was 308 K, and the fraction of thermal neutrons was approximately 95% at the reference position of the thermal neutron field. The thermal neutron fluence rate was determined by adopting the Westcott convention method based on neutron activation analysis using a gold foil. The Westcott fluence rate for thermal neutron at the reference position was (2326.7 ± 8.4) cm−2s−1. The true thermal neutron fluence rate at the reference position in the KRISS thermal neutron field was (2700 ± 29) cm−2s−1 (at the reference date of June 30, 2014). The response of a spherical proportional counter with He-3 (SP9 neutron detector) was evaluated in the newly established field as (3.083 ± 0.045) cm2 for the reference calibration condition (a parallel neutron beam with a Maxwellian energy distribution having a most probable energy of 0.025 eV).

LiCaAlF6 scintillators in neutron and gamma radiation fields

Intentionally doped LiCaAlF6 (LiCAF) single crystals are prospective scintillators, especially for thermal neutron detection through the 6Li(n,t)4He nuclear reaction. Four different LiCAF scintillator samples were tested in various neutron and gamma fields. Two of the tested samples were LiCAF:Eu and LiCAF:Eu,Na single crystals, and another two samples were made of LiCAF:Eu micro crystals dispersed in transparent rubber, with different rubber dimensions. All LiCAF samples contain lithium enriched to[Formula: see text]Li. A plutonium–beryllium source was used as a neutron source. The neutron spectrum was modified by moderator and filter to get different ratios between thermal, epithermal and fast neutron fluence rates. The MCNP code was used for calculations of the fluence rates for different configurations. Radionuclides [Formula: see text]Cs and [Formula: see text]Co were applied as gamma radiation sources. The light signal from the scintillator was evaluated with a photomultiplier and a multichannel analyzer. The purpose of this work was to study the characteristics of LiCAF scintillators, especially the ability to discriminate signals from neutron and gamma radiation, which is the basic scintillator condition for neutron detection in mixed neutron-gamma radiation fields. Generally, the discrimination can be done by the pulse height and/or the pulse shape of the evaluated signals. Both methods can be used for a LiCAF scintillator. However, only the pulse height discrimination method is discussed in this paper. The possibility of fast neutron detection with LiCAF scintillators was also tested.

Photo neutron dose equivalent rate in 15 MV X-ray beam from a Siemens Primus Linac.

Fast and thermal neutron fluence rates from a 15 MV X-ray beams of a Siemens Primus Linac were measured using bare and moderated BF3 proportional counter inside the treatment room at different locations. Fluence rate values were converted to dose equivalent rate (DER) utilizing conversion factors of American Association of Physicist in Medicine's (AAPM) report number 19. For thermal neutrons, maximum and minimum DERs were 3.46 × 10-6 (3 m from isocenter in +Y direction, 0 × 0 field size) and 8.36 × 10-8 Sv/min (in maze, 40 × 40 field size), respectively. For fast neutrons, maximum DERs using 9” and 3” moderators were 1.6 × 10-5 and 1.74 × 10-5 Sv/min (2 m from isocenter in +Y direction, 0 × 0 field size), respectively. By changing the field size, the variation in thermal neutron DER was more than the fast neutron DER and the changes in fast neutron DER were not significant in the bunker except inside the radiation field. This study showed that at all points and distances, by decreasing field size of the beam, thermal and fast neutron DER increases and the number of thermal neutrons is more than fast neutrons.

Applicability of self-activation of an NaI scintillator for measurement of photo-neutrons around a high-energy X-ray radiotherapy machine.

The applicability of the activation of an NaI scintillator for neutron monitoring at a clinical linac was investigated experimentally. Thermal neutron fluence rates are derived by measurement of the I-128 activity generated in an NaI scintillator irradiated by neutrons; β-rays from I-128 are detected efficiently by the NaI scintillator. In order to verify the validity of this method for neutron measurement, we irradiated an NaI scintillator at a research reactor, and the neutron fluence rate was estimated. The method was then applied to neutron measurement at a 10-MV linac (Varian Clinac 21EX), and the neutron fluence rate was estimated at the isocenter and at 30cm from the isocenter. When the scintillator was irradiated directly by high-energy X-rays, the production of I-126 was observed due to photo-nuclear reactions, in addition to the generation of I-128 and Na-24. From the results obtained by these measurements, it was found that the neutron measurement by activation of an NaI scintillator has a great advantage in estimates of a low neutron fluence rate by use of a quick measurement following a short-time irradiation. Also, the future application of this method to quasi real-time monitoring of neutrons during patient treatments at a radiotherapy facility is discussed, as well as the method of evaluation of the neutron dose.

Solid state detectors for dosimetry in the BNCT beam of the LVR-15 research reactor

Dosimetry measurements have been carried out at the epithermal neutron beam of LVR-15 reactor (Řež) moderated by a polyethylene disk placed through the collimator exit. The spatial distributions of thermal neutron fluence rate and of gamma and fast neutron dose rates have been attained. The contributions to the dose of the radiation field components have been assessed with Fricke-gel-layers, by pixel-to-pixel elaboration of the light transmittance images of samples having different isotopic composition and with thermoluminescence detectors, by analysing the shape of the glow curve. An analogous experiment has been carried out at the thermal column HK1 of the same LVR-15 reactor, aimed at neutronography research. Good consistency of all results was obtained.

New PTB thermal neutron calibration facility: first results

A new thermal neutron calibration facility based on a moderator assembly has been set up at PTB. It consists of 16 (241)Am-Be radionuclide sources mounted in a graphite block, 1.5 m wide, 1.5 m high and 1.8 m deep. The sources are distributed to eight different positions, at a mean distance of ∼1.25 m from the front face of the moderator. The neutron field at the reference position, 30 cm in front of the moderator device and 75 cm above the floor, has been characterised using calculations, Bonner sphere measurements and gold foil activation. First results are shown. The field is highly thermalised: 99 % in terms of fluence. It is quite homogenous within a 20 cm×20 cm area, but the absolute value of the thermal neutron fluence rate is small and yields an ambient dose equivalent rate of 3 µSv h(-1).

A new active thermal neutron detector

This communication presents the main results about the design and in-house fabrication of a new solid-state neutron detector, which produces a DC output signal proportional to the thermal neutron fluence rate. The detector has been developed within the framework of the 3-y project NESCOFI@BTF of INFN (CSN V). Due to its sensitivity, photon rejection, low cost and minimum size, this device is suited to be used in moderator-based spectrometers.

Moderator for neutron activation with the photoneutrons produced by a LINAC

Using Monte Carlo methods a polyethylene moderator has been designed to induce activation using the photoneutrons field of a 15 MV linear accelerator for radiotherapy. In the calculations the photoneutron spectrum at 1 m from the isocenter was used as a source term and the neutron spectra were calculated in the center of different size cylindrical moderators. The best size was selected defining the thermal-to-fast-neutron ratios as a figure of merit. The moderator was built and its performance was evaluated by inducing the activation of Mn dissolved in water, silver coins and souvenir coin. The thermal neutron fluence rate was determined with the Mn samples being 9.96 × 105 cm−2 Gyx−1.

Comparing neutronics codes performance in analyzing a fresh-fuelled research reactor core

In this paper the relative performance of different simulation approaches is examined, focusing on the neutron fluence rate distribution in a nuclear reactor core. The main scope of the work is to benchmark and validate the neutronics code systems utilized in the Greek Research Reactor (GRR-1) for a high-density Low Enriched Uranium (LEU) core of compact size. For this purpose the recently converted core of the Portuguese Research Reactor (RPI), fueled with fresh, low enrichment in U-235 fuel, was simulated with the stochastic code TRIPOLI and the deterministic code system XSDRN/CITATION. RPI was selected on the basis that it is a similar to GRR-1 pool-type reactor, using same fuel and control rods type, as well as same types of coolant, moderator and reflector. The neutron fluence rate in RPI was computed using each numerical approach with changed approximations. In this frame the stochastic code TRIPOLI was tested using two different nuclear data libraries, i.e., ENDF/B-VI versus JEFF3.1, and two different ways of source definition, i.e., “point sources”, placed in the center of each fuel cell, versus a “distributed source”, where each fuel volume was considered as a neutron source. The deterministic code system XSDRN/CITATION was tested with respect to the definition of the transverse leakages associated to each one-dimensional, user-defined core zone, as analyzed by the XSDRN code in order to provide the zone equivalent cross sections. Thermal, epithermal and fast neutron fluence rates were computed and local values found in a 15cm segment immediately below the fuel mid-height were compared to activation foil measurements, as well as to corresponding MCNP results obtained at the RPI. The comparisons were performed in representative core positions, including standard fuel assemblies, dummy (non-fuelled) assemblies, beryllium reflectors and free grid positions close to the core. Application options and future improvements of the tested codes are discussed. Finally it is worth noting that this paper, including measurements and calculations by three different codes and tests of different cross section libraries for a new commissioned core, can provide useful material for benchmarking neutronics codes that are under development in various reactor laboratories, as well for optimizing codes already in use.

Usability of epoxy/ilmenite composite material as an attenuator for radiation and a restoration mortar for cracks

In the study, the usability of epoxy/ilmenite (EP/IIm) composite material as a radiation shielding in many applications and a restoration/injection mortar for cracks developing in biological concrete shields was investigated. Radiation attenuation properties of EP/IIm composite material was searched out using 252Cf (100μg) neutron source and gamma spectrometer with stilbene organic scintillator. Fast neutron fluence rates and gamma fluxes were measured and displayed as energy distributions and attenuation relations. Thermal neutron fluence rates were measured using BF3 detector and displayed as attenuation relations. The experimental attenuation parameters; macroscopic effective removal cross-section ΣR (cm−1), total attenuation coefficient μ (cm−1) and macroscopic cross-section Σ (cm−1) of fast neutrons, gamma rays and thermal neutron respectively have been determined. Theoretical calculations have been achieved using MCNP-4C2 code to evaluate the concerned parameters. Also, MERCSF – N program had been used to recalculate ΣR (cm−1). The obtained experimental and theoretical results were compared, where a reasonable agreement was recognized.

Optimisation of the body composition analyser using the gamma shield and improvement of accuracy using ORNL anthropomorphic phantoms

The analysis of body elements using the prompt gamma rays neutron activation method is a very useful and highly accurate method Lone, M. A., Mughabghab, S. F. and Paviotti-Corcuera, R. Development of a database for prompt gamma-ray neutron activation analysis. Summary Report of the Second Research Coordination Meeting. IAEA Headquarters. IAEA Nuclear Data Section. INDC(NDS)-424. (2001) that has many applications in different fields such as the diagnosis of specific diseases, including certain types of cancers. To protect patients from gamma rays produced by the system, an efficient strategy is to apply a gamma shield. In this study, the gamma shield was placed in three separate positions. The influence of these positions on reducing the effective dose was examined in a 5-y-old Oak Ridge National Laboratory mathematical phantom. Other parameters considered were sensitivity and coefficient of variation (CV) of thermal neutron fluence rate. With the best configuration, the total effective dose per minute (ET) was decreased ∼52 % and the sensitivity was ∼2.03-fold higher than when no shield was present.

A new monitor set for the determination of neutron flux parameters in short-time k0-NAA

Multipurpose research reactors such as LVR-15 in Řež require monitoring of the neutron flux parameters (f, α) in each batch of samples analyzed when k0 standardization in NAA is to be used. The above parameters may change quite unpredictably, because experiments in channels adjacent to those used for NAA require an adjustment of the reactor operation parameters and/or active core configuration. For frequent monitoring of the neutron flux parameters the bare multi-monitor method is very convenient. The well-known Au–Zr tri-isotopic monitor set that provides a good tool for determining f and α after long-time irradiation is not optimal in case of short-time irradiation because only a low activity of the 95Zr radionuclide is formed. Therefore, several elements forming radionuclides with suitable half-lives and Q0 and Ēr parameters in a wide range of values were tested, namely 198Au, 56Mn, 88Rb, 128I, 139Ba, and 239U. As a result, an optimal mixture was selected consisting of Au, Mn, and Rb to form a well suited monitor set for irradiation at a thermal neutron fluence rate of 3×1017 m−2 s−1. The procedure of short-time INAA with the new monitor set for k0 standardization was successfully validated using the synthetic reference material SMELS 1 and several matrix reference materials (RMs) representing matrices of sample types frequently analyzed in our laboratory. The results were obtained using the Kayzero for Windows program.

Comparison of the neutron energy spectrum and neutron fluence rate in the LVR-15 research reactor with fuel with different enrichment

This contribution compares measured neutron energy spectra and neutron fluence rates in the LVR-15 reactor core fully equipped with IRT2M nuclear fuel (enrichment 36% of 235U in the form of UO2) and then with a partially replaced core equipped with three IRT4M nuclear fuel assemblies (enrichment 19.7%). The measurements were performed in the LVR-15 reactor at Research Center Rez Ltd. in the Czech Republic, and were related to a planned transition to low-enriched nuclear fuel within the scope of the RERTR programme. An activation method was chosen for the neutron spectrum measurement. Iron, cobalt, nickel, copper, titanium, iridium and niobium foils were irradiated at four positions near the replaced fuel assemblies. Reaction rates for observed reaction channels were determined using gamma spectroscopy. Reaction rates along the height of the reactor core at the same positions were determined using iron, nickel, and cobalt foils. The SAND-II and STAYNL computer programs were used for neutron spectrum adjustment, and input approximation for both programs was calculated using MCNPX (v2.6). The results include a comparison of theoretical and measured data. Differences were found between thermal neutron fluence rates inside IRT2M fuel assemblies and IRT4M fuel. This difference was predicted by preliminary calculations, but it becomes less significant as distance from fuel assemblies increases.