252Cf Neutron Source Research Articles

Calculation of response parameters for a neutron long counter instrument

The primary goal of this study was to develop a simulation model of a long counter available at Canadian Nuclear Laboratories (CNL). Using the Monte Carlo N-Particle version 6 (MCNP6) code, the model was used to calculate, as a function of incident energy, the number of counts recorded per source neutron, effective centre, and sensitivity. This study also carried out measurements of the neutron emission rate of and direct neutron flux at 2 m from an in-house 252Cf neutron source.

Optimizing gamma-ray shielding for boron neutron capture therapy by using unglazed ceramic composites

In this study, the researchers investigated the effectiveness of different ceramic samples in filtering gamma rays emitted by various sources. The samples, which contained varying concentrations of lead and barium, were evaluated for their potential use in beam-shapingassembly systems for Boron Neutron Capture Therapy techniques, specifically those using the 252Cf neutron source for cancer treatment. Stilbene and NaI (Tl) were used to measure the total and pure gamma rays in this study. In addition, theoretical calculations using the Phy-X/PSD software have been conducted across a broad energy range, spanning from 0.001 to 18 MeV. These calculations aim to determine the attenuation of gamma rays and the relative deviations compared to the corresponding measured gamma-ray energies. Various composites with different percentages of additives and thicknesses, ranging from 0.8 cm to 4 cm, have shown satisfactory attenuation properties for gamma-ray shielding applications. However, the samples containing lead, especially the unglazed ceramic composites with 15 % lead and 4 cm thickness, were found to have the most preferred attenuation properties. The results obtained indicate that samples with a higher lead content are more effective attenuators compared to those with additional barium. In addition, the relationship between the transmission gamma ray factor and the percentage of additives and energies of gamma rays can be determined from the experimental linear attenuation coefficient coefficients. In addition, various coefficients such as the mass attenuation coefficient, half-value layer, effective atomic numbers, and effective electron densities are studied theoretically for the unglazed ceramic samples. The dose rate is also taken into consideration.

Monte Carlo simulation and optimization of neutron ray shielding performance of related materials

Polymers have become widely used substrate materials for shielding neutron rays because of their high hydrogen content and easy processing procedures. Rare-earth materials are also being gradually adopted as neutron absorbers because of their considerable thermal neutron absorption cross-sections. This paper utilizes the FLUKA Monte Carlo simulation program to compare the shielding effects of various polymers and rare-earth oxides on neutron rays across different energy ranges. The study investigates the superior shielding materials for neutron radiation in each energy range. Subsequently, a series of materials are simulated by combining the preferred shielding materials for neutron rays in each energy range, exploring the influence of material composition and composite structure on the effectiveness of neutron ray shielding. It is revealed that the preferred material for shielding neutron rays changes for different energy ranges. For low-energy neutron rays, rare-earth oxides such as Sm2O3 and Gd2O3 demonstrate the most effective shielding, whereas for high-energy neutron rays, polyethylene (PE) provides the best shielding performance. Materials with different compositions show varying preferred structures when dealing with a 252Cf neutron source. However, in mitigating the secondary gamma rays generated during the neutron shielding process, stacked-type materials exhibit the most effective shielding performance.

Sm2O3 micron plates/B4C/HDPE composites containing high specific surface area fillers for neutron and gamma-ray complex radiation shielding

In this study, a series of Sm2O3 micron plates/B4C/HDPE composites composed of synthesized Sm2O3 fillers (Sm2O3 micron plates) are prepared for shielding neutron and gamma radiation. The influence of micromorphology of Sm2O3 fillers on neutron and gamma radiation shielding properties of composites is investigated in detail. The XRD pattern reveals that the phase of synthetic Sm2O3 is cubic crystal systems, body-centered cubic lattices, and its space group is Ia 3‾ (206). SEM images and BET analyses reveal that the micromorphology of synthesized Sm2O3 is micron plates. The BET-specific surface area of the Sm2O3 fillers is increased with addition of urea content. The differential scanning calorimetry (DSC) curves reveal that Sm2O3 fillers increase the melting temperature of the composites, which is up to 138.6 °C. The thermogravimetric analysis (TGA) results reveal that the initial thermal degradation temperatures of the composites are all above 440 °C. The neutron and gamma radiation shielding tests show that Sm2O3 fillers with high BET-specific surface area (8.20 m2/g) and uniform size improve the neutron and gamma shielding rate of composites. A superior composite containing 10 wt% Sm2O3 (R = 1:25, R value represents the molar ratio of rare earth elements to urea)/20 wt% B4C/70 wt% HDPE has a neutron radiation shielding rate of 98.7% with a thickness of 15 cm under the 252Cf neutron source and a gamma radiation shielding rate of 72.1% with a thickness of 15 cm under 137Cs gamma source. And these lead-free and environment-friendly composites can be widely used in the neutron and gamma complex radiation fields.

Optimization of pulse shape discrimination based on frequency domain for SiPM-based NaIL scintillator detector

Frequency gradient analysis (FGA) neutron-gamma pulse shape discrimination (PSD) based on Fourier transform has proved to be an effective method for discrimination between two types of pulses by simply constructing a discrimination parameter by subtracting the amplitude of the 1st frequency component from the 0th frequency component. However, the discrimination parameter constructed by the above method could not meet the requirements at some lower sample rates, and the dark noise caused using optoelectronic devices such as miniaturized SiPMs array further erodes the difference between the two frequency components. This paper proposes a new discrimination parameter construction method based on FGA for neutron-gamma discrimination, which is referred as mFGA (modified FGA). This discrimination parameter uses the 0th frequency component and other frequency components to enhance the discrimination performance. A neutron detector module consisting of a 25.4 mm× 25.4 mm× 50.8 mm NaI:Tl,6Li (NaIL) scintillator and an array of 8 × 8 SiPMs was established to verify the method of mFGA. The module measured a 252Cf neutron source under three shielding conditions: no shield, 5 cm high-density polyethylene (HDPE), and 10 cm lead brick. The proposed mFGA method was used to calculate the figure of merit (FoM) under each shielding condition and compared to the traditional frequency gradient method (FGA) and the traditional charge comparison method (CCM). The results show that the discrimination performance with the mFGA method is significantly improved. Taking the HDPE shielding group as an example, the performances of the three methods are FoM-FGA =1.14, FoM-mFGA =2.75, FoM-CCM =0.92. At the same time, with the increase of energy, due to the neutron gamma pulse becoming similar, and the discrimination performance of FGA and CCM decrease significantly. The discrimination performance of the mFGA method is relatively stable in all energy regions. Therefore, the proposed method is superior to the traditional FGA of the 0th frequency minus the 1st frequency and has a good application prospect for neutron-gamma discrimination with detectors of poor signal quality.

Development of a neutron yield measurement system utilizing BF_3 detectors on the HL-3 tokamak

A neutron yield measurement system, comprising five BF3 proportional counters with excellent n/γ discrimination capabilities, has been meticulously developed and implemented on the HL-3 tokamak to provide time-resolved assessments of neutron emission rates. The size of moderator, optimized for the flat energy responses, was skillfully simulated and designed using the MCNP code. An in-depth analysis of measurement errors stemming from pile-up phenomena determined that the upper limit of the count rate, under the current electronics design, is 131 kcps with a relative error of less than 20%. Drawing from the pulse amplitude spectrum obtained from a 252Cf neutron source and experimentation on the HL-3 tokamak, the amplitude threshold for the BF3 neutron detector systems was judiciously established. Furthermore, it was conclusively affirmed that the neutron yield measurement system possesses the capability to effectively discriminate between neutron and gamma signals. During HL-3 discharges, deuterium-deuterium (DD) neutrons were reliably detected with a temporal resolution of 10 ms, and the variations in neutron count rates was found to be in alignment with the neutral beam injection. Ultimately, the total neutron yield was successfully estimated through the utilization of count rates from four BF3 detectors, thereby validating the robust performance of the neutron yield measurement system.

The contribution of the air and room scattered neutrons in the PTB calibration facility

Neutron detectors and dosimeters used for radiation protection purposes are commonly calibrated with radioisotope-based neutron sources in dedicated calibration facilities. Due to interactions of the neutrons emitted by the sources with components of the calibration room, the neutron fluence at the measuring point is a sum of several components. As the results of calibration procedures should ideally be independent of the environment in which the measurements are performed, the effects caused by neutrons which are not directly reaching the device under test need to be considered, i.e. the influence of scattered neutrons needs to be determined and corrections have to be made. The contribution of the neutrons scattered from the air and the room to the fluence and dose must be taken into consideration, especially in the case of small and medium sized calibration rooms. The scattered neutron component can be evaluated numerically provided a detailed model of the calibration facility is available. In this work, a realistic MCNP6 model of the calibration room at the Physikalisch-Technische Bundesanstalt (PTB) has been developed in order to evaluate carefully the contribution of scattered neutrons from the air, wall, and equipment in the irradiation room for 252Cf, 241Am–Be and D2O moderated 252Cf (with and without Cd-shielding) neutron sources at different reference points and with and without shadow objects. The calculation also considers new fission spectral data for the neutron reference field and recent cross section data libraries. The results show that the contribution of air out-scattered neutrons depends on the detector position and on the neutron source spectrum. This effect is particularly significant for the moderated 252Cf source. More than a third of the total scored neutrons come from the walls and ceiling and the total contributions of the scattered neutrons to the total fluence at the reference position 170 cm away from the source are 50.6 %, 46.6 %, 55.0 % and 52.5 % for 252Cf, 241Am–Be, D2O-252Cf–Cd and D2O-252Cf sources respectively.

Feasibility of using BeO rods as secondary neutron sources in the long-life fuel cycle high-temperature gas-cooled reactor

External sources of neutron provide stable and sufficient neutron for initial startup of a nuclear reactor. They also provide signals for neutron detectors to monitor the safety of reactor during shutdown. In the high temperature engineering test reactor (HTTR), 252Cf is used as the external neutron source. However, the 252Cf sources must be renewed every approximately 7 years because of its relatively short half-life of 2.6 years. The renewal of 252Cf sources requires a high cost and a very complicated procedure. This study investigated the feasibility of using BeO rods as the secondary neutron sources to avoid renewing the 252Cf neutron sources periodically. The BeO rods could exist in the reactor for a long time so that if the reactor operates long enough, the neutron flux at the wide-range monitoring detectors remains more than 10n.s−1.cm−2 even if the reactor is shutdown for as long as 5 years. The results of this study indicated that using BeO rods as the secondary neutron sources would be an attractive option for the future HTGR design with a long-life fuel cycle.

Neutron spectroscopy with TENIS using an artificial neural network

In this research, a ThErmal Neutron Imaging System (TENIS) consisting of two perpendicular sets of plastic scintillator arrays for boron neutron capture therapy (BNCT) application has been investigated in a completely different approach for neutron energy spectrum unfolding. TENIS provides a thermal neutron map based on the detection of 2.22 MeV gamma-rays resulting from 1H(nth, γ)2D reactions, but in the present study, the 70-pixel thermal neutron images have been used as input data for unfolding the energy spectrum of incident neutrons. Having generated the thermal neutron images for 109 incident mono-energetic neutrons, a 70 × 109 response matrix has been generated using the MCNPX2.6 code for feeding into the artificial neural network tools of MATLAB. The errors of the final results for mono-energetic neutron sources are less than 10% and the root mean square error (RMSE) for the unfolded neutron spectrum of 252Cf is about 0.01. The agreement of the unfolding results for mono-energetic and 252Cf neutron sources confirms the performance of the TENIS system as a neutron spectrometer.

A method of obtaining the neutron multiplicity distribution online based on the list-mode data acquisition

The obtainment of the neutron multiplicity distribution is the precondition of the neutron coincidence or multiplicity measurement. In related measurements, the calibration of the detector and the optimization of the analysis parameters prefer acquiring the neutron multiplicity distribution through the offline time-correlation analysis of the measured time stamps, while the usual measurement tasks tend to get it online. To meet with the demands of online and offline obtainment of the neutron multiplicity distribution, an online time-correlation analysis method on the basis of list-mode data acquisition technique is proposed in present work. The reliability of the obtained neutron multiplicity distribution is verified through the experiments carried out with a strong and a weak 252Cf neutron sources.

Benchmarking Neutron Counting System for Passive Measurements and Active Interrogation of Unknown Objects for Fissile and Fissionable Materials Determination

The low rate at which some fissile isotopes, such as 235U, 233U, and 239Pu, undergo spontaneous fission leads to a weak signal, resulting in a high-uncertainty in applying passive neutron counting techniques. Stimulating fission through active neutron interrogation can overcome this issue. At Canadian Nuclear Laboratories, a 252Cf and a deuterium-deuterium neutron source are available. In this study, a neutron counting system was designed to perform passive measurements and active neutron interrogation for a search of special nuclear material. The detection system consists of a cylindrical cavity surrounded by a polyethylene moderator with 3He detectors interspersed throughout. When used for passive measurements, the sample is placed in the cylindrical cavity, whereas in active interrogation mode, the 252Cf neutron source and the sample are placed in close proximity to each other in the cylindrical cavity. Measurements that actively interrogated samples, notably containing (among other isotopes) either 235U or 239Pu whose mass was on the order of fractions of a gram, carried out using the 252Cf neutron source found that the average delayed neutron count rate was on the same order of magnitude as those obtained from passive measurements using several kilograms of natural uranium. The Monte Carlo N-Particle 6 version 2.0 radiation transport code was used to simulate the aforementioned active interrogations and to inform the experimental results. Results showed that, due to the close proximity of the polyethylene moderator to the 252Cf source, the neutron energy spectrum traversing the fissile sample has a significant thermal component that maximizes the fission reaction rate in the interrogated fissile samples, thereby allowing for successful measurements.

Time-dependent multiplicity for Cf-252 neutron source

A Cf-252 neutron source is widely used for testing and calibrating neutron multiplicity counters. General equations for calculating the time-dependent strength and multiplicity of Cf-252 sources are deduced based upon the decay model of Cf-252, Cf-250, and their daughter products Cm-248 and Cm-246. With the nuclear data of these four nuclides, an instance of a long-aged (>40 a) Cf-252 source is presented to illustrate the variation of the strength and multiplicity with time; the calculation results indicate that the 1st, 2nd, and 3rd moment factorials of the neutron multiplicity are significant decreased compared to the ones of nuclide Cf-252. For verification, a neutron multiplicity counting experiment for this Cf-252 source (I#) and another Cf-252 source (II#) with 17.1-year service life was carried out respectively using a thermal neutron multiplicity counter. The measured results are consistent with the calculation results from equations. The results from this study help to understand the changes in attributes with respect to time for any Cf-252 source while making appropriate corrections for obtaining accurate calibration results.

Effects of Neutron and Gamma Rays Combined Irradiation on the Transcriptional Profile of Human Peripheral Blood.

We studied the effects of neutrons, neutrons and γ rays, and γ rays exposures on the transcription spectrum in human peripheral blood of three healthy adult men. Samples were irradiated with 1.42 Gy 2.5-MeV neutrons, 0.71 Gy neutrons and 0.71 Gy 137Cs γ rays, and 1.42 Gy 137Cs γ rays. Transcriptome sequencing identified 56 differentially co-expressed genes and enriched 26 KEGG pathways. There are 97, 45 and 30 differentially expressed genes in neutron, neutron and γ ray combined treatment, and γ rays, respectively, and 21, 3 and 8 KEGG pathways with significant differences are enriched. Fluorescence quantitative polymerase chain reaction (qPCR) verified differential co-expression of AEN, BAX, DDB2, FDXR, and MDM2. Additionally, irradiation of AHH-1 human lymphocytes with a 252Cf neutron source at 0, 0.14, 0.35, and 0.71 Gy, fluorescence qPCR revealed a dose-response relationship for BAX, DDB2, and FDXR at dose ranges of 0-0.71 Gy, with R2 of 0.803, 0.999, and 0.999, respectively. Thus, neutrons can induce more differentially expressed genes and enrich more pathways. Combined treatment of neutrons and γ-rays may incorporate damage of both high and low LET, the genes activated by neutrons and γ rays combined are almost the combination of genes activated by neutron and γ rays combined treatment. BAX, DDB2 and FDXR are differentially expressed after irradiation by Deuterium-Deuterium (D-D) neutron source and 252Cf neutron source, so they are expected to be molecular targets of neutron damage.

THE IMPACT OF NEW ICRU95 QUANTITIES AND SPECTRUM DATA ON THE DOSIMETRIC QUANTITIES OF THE NEUTRON REFERENCE FIELDS AT PTB.

In 2020, the ICRU released a new report which includes the re-definition of the operational quantities used in radiation protection and new conversion coefficients from physical quantities to operational radiation protection quantities. An assessment of the ambient and personal dose conversion coefficients for the reference neutron fields of radionuclide sources at PTB is necessary based on these new definitions. In this work, a numerical estimation of the conversion coefficients of moderated and unmoderated 252Cf and 241Am-Be neutron sources based on ICRU57 and ICRU95 reports and using spectrum data available in the ISO 8529-1 standard and at PTB are discussed. Two numerical approaches are used for this estimation to ensure the reliability of the calculated values: a direct calculation using MCNP6, and cubic interpolation of conversion coefficients datasets written in Python. The results show large differences between the spectrum-averaged operational quantities for the current and new conversion coefficients of up to 23%. The choice of spectrum data affects conversion coefficient values by 6-8%.

Estimating 10B and 14N doses using a pair of LiF-based TLD-600 and TLD-700 as an alternative to gold activation method

Purpose:A new method using TLD-600 and TLD-700 was proposed to measure a thermal neutron flux, which is used to evaluate 10B and 14N doses for boron neutron capture therapy (BNCT). Material and method:TLD-600 and TLD-700 were enriched by 6Li (95.6%) and 7Li (99.99%), respectively. The TLDs were calibrated using a 252Cf neutron source within a 30 cm-diameter D2O sphere to produce a thermal neutron flux calibration factor. The calibration factor was obtained by dividing the neutron-induced TLD readings by the delivered thermal neutron fluence. In addition, a neutron self-shielding correction was calculated to compensate for the perturbed neutron fluence between adjacent TLDs. To validate the procedure involved in this new method, TLDs were irradiated by thermal neutron fluxes from the research/educational reactor. The results were compared with the thermal neutron fluxes by the conventional gold activation method. Results and discussion:The calibration factor from TLDs irradiated by the thermal neutron flux from the 252Cf neutrons was determined 3.80 × 104 cm−2 nC−1. Using this factor, the thermal neuron flux at the reactor was measured 2.28 × 105 cm−2 W−1 s−1 by the TLD method. The one measured by the gold activation method was 2.17 × 105 cm−2 W−1 s−1 showing a difference of 5%. With the thermal neutron fluence measured by TLDs, a Li-6 dose of TLD-600 (Dn,αTLD) can be calculated. Then the two major components of BNCT dosimetry, a B-10 dose (D(n,α)B−10) and an N-14 dose (D(n,p)N−14) can be calculated by multiplying Dn,αTLD with constants due to a strong 1/v behavior of the three reactions. Conclusions:The TLD method developed in this study was successfully validated against an unknown mixed neutron/gamma field from the reactor, showing about 5% difference in measured thermal neutron fluences when compared to the conventional gold activation method. Therefore, we concluded that the two major dose components of BNCT, 10B and 14N doses can be accurately evaluated by the neutron-induced TLD dose.

Accurate activity determination of a californium neutron source

Accurately counting analog events requires constructing an electronics chain that produces one count for each input pulse. In this work we review the use of Nuclear Instrumentation Module electronic units for counting neutron capture events in a 3He tube. We identify two unique types of false trigger events in a leading-edge discriminator and show how a dual timer module can be used to produce a veto window to exclude these events. We use the constructed electronics chain to build an apparatus to measure neutron pulses from a 252Cf neutron source. We compare the measurements with a Monte Carlo N-Particle (MCNP) model to determine the activity of the neutron source. Furthermore, by making additional measurements with borated polyethylene attenuators between the source and detector, we are able to determine the boron concentration of the polyethylene. This technique provides accurate determination of the source activity to a precision of 2.8% at the k = 1 level. The method used is simple, inexpensive, and requires no additional calibrated instruments.

Development of the neutron salt-meter RANS-μ for non-destructive inspection of concrete structure at on-site use

We have started the development of a transportable neutron salt-meter, we call it RANS-μ, combining a 252Cf neutron source and a prompt gamma neutron activation analysis. Trials of chloride detection measurement with RANS-μ were performed outdoor using removed bridges damaged by chloride attack at an outdoor yard in Public Works Research Institute (PWRI) and a test bridge in Fukushima Robot Test Field. For the measurement at PWRI, the results obtained by RANS-μ were compared with those of the automatic potentiometric titration by drilled powder, and then consistent results were obtained.

Gamma and neutron separation using emission wavelengths in Eu:LiCaI scintillators

Abstract Scintillators have long been known as radiation detectors and are still used in various applications. Recently, scintillators containing 6Li have been developed as neutron detectors and have attracted attention. 6Li absorbs thermal neutrons and emits α+3H, which is promising as a neutron detector if it can be separated from background gamma rays. We have been developing Eu:LiI–CaI2-based scintillators (Eu:LiCaI) for this purpose. In scintillator detectors, waveform information is generally used to distinguish particles such as neutrons and gamma rays. We propose a new particle identification method using emission wavelength information. In this study, experiments were conducted using Eu:LiCaI crystals, multi-pixel photon counter optical sensors, and long-wavelength cut filters to verify the proposed method. The results of irradiating a 252Cf neutron source and a 60Co gamma-ray source indicate that there is a particle dependence on the output signal ratio with and without filters. This indicates that different types of radiation particles have different emission wavelengths. This is the first demonstration of a wavelength-based particle identification method.

Use of nickel sphere and copper cube with 252Cf neutron source in the centre for test of nuclear data libraries ENDF/B-VII.1, ENDF/B-VIII.0, JEFF- 3.3, BROND- 3.1

The leakage neutron spectrum measurements have been done on benchmark spherical assembly-nickel sphere with a diameter of 50 cm and a copper cube (block) with dimensions of 49.5 x 49.5 x 48 cm3 in Research Centre Rez (RC Rez). The 252Cf neutron source was placed into the centre of nickel sphere and copper cube. The proton recoil method was used for the neutron spectrum measurement using spherical hydrogen proportional detectors (HPD) with pressure of 400 and 1000 kPa (diameter of detectors is 4 cm) and scintillation stilbene (ST) detector (diam. of 1 x 1 cm). The neutron energy range of spectrometer is from 0.04 MeV to 1.3 MeV for HPD and from 1 MeV to 12 MeV for ST. The adequate MCNP neutron spectrum calculations based on data libraries ENDF/B-VII.1, ENDF/B-VIII.0, JEFF-3.3, BROND-3.1 were done and compared with the experiment, i.e., calculation to experiment ratio C/E was determined.

Progress in the Evaluation and Validation of n+56,57Fe Cross Sections

There has been a continued effort since 2019 within the IAEA INDEN collaboration to improve the evaluation of neutron induced reactions on iron isotopes. The reason for the 30% underestimation of the neutron leakage spectrum from a thick iron sphere was found primarily to be due to the overestimation of the inelastic cross sections in the 56Fe evaluated data file produced within the CIELO project of the OECD/NEA Data Bank. The over-estimation of the neutron flux between the resonances near 300 keV was traced to neglecting the fluctuating nature of the total cross section of 57Fe in the fast neutron energy range, since the evaluated resolved resonance range of 57Fe extended only up to 190 keV. The added 1/v background in the "iron window" below 28 keV is in excellent agreement with the independently evaluated one in the JENDL-5.0 library that included the direct capture component in the evaluation. Performance of the updated 56,57Fe evaluations was tested on a set of criticality benchmarks from the ICSBEP Handbook, including the dependence on reflector thickness and on new deep penetration shielding benchmark using a 252Cf(sf) neutron source undertaken at Rez, Czech Republic. Neutron leakage for 43 MeV incident neutrons was also validated.