Moderated 252Cf Source Research Articles

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.

A feasibility study of a H2O-moderated 252Cf source for metrology

The new edition of ISO 8529 series reduced from 4 to 3 standard neutron reference field. Recent studies have highlighted the increasing importance of characterizing neutron source spectra at different energies to obtain appropriate fluence-to-dose-equivalent conversion coefficient. This study presents the determination of neutron energy fluency from a free and moderated 252Cf (120 µg) source placed within spheres of different diameters containing distilled water. The spectrum determination is based on both Monte Carlo simulations and experimental measurements with Bonner Sphere Spectrometer (BSS) of multiple diameters: 5.08 cm (2"), 7.62 cm (3”), 12.70 cm (5”), 20.32 cm (8”), 25.40 cm (10”), and 30.48 cm (12”). The aim was to characterize the new water-moderated spectra, obtaining reference values for operational quantities that represent realistic radiation fields for testing, calibrating, and irradiation of personal dosimetry and survey meters.

CALCULATIONS FOR THE HEAVY WATER MODERATED 252CF REFERENCE FIELDS AT PTB.

Knowledge of the neutron spectra of reference fields is very important for calibration purposes. This paper presents new calculations of the neutron spectra and dosimetric parameters of a D2O moderated 252Cf source assembly used to produce a neutron reference field at the Physikalisch-Technische Bundeanstalt (PTB), based on recent 252Cf fission neutron spectrum data and conversion coefficients for operational radiation protection units recommended by the International Commission on Radiological Protection (ICRP) report 57. Neutron fluence and dose calculations were carried out using a detailed simulation model of the source assembly implemented using MCNP6 and the cross-section data library ENDF/B-VIII.0. The PTB moderator was compared with the International Organization for Standardization, ISO 8529 model and to previous calculations done at PTB. The results show differences to the previous calculations performed at PTB of the order of 5% and a slight discrepancy with the ISO 8529 moderator model regarding the values of the operational quantities.

Pulse-Shape Discrimination of SiPM Array-Coupled CLYC Detector Using Convolutional Neural Network

Cs2LiYCl6: Ce3+ (CLYC) is a dual-mode gamma-neutron scintillator with a medium gamma-ray resolution and pulse-shape discrimination (PSD) capability. The PSD performance of CLYC is greatly weakened when coupled with silicon photomultipliers (SiPMs) because of SiPMs’ low detection efficiency for the ultrafast Core-Valence-Luminescence (CVL) component under gamma excitation. In our previous work, the PSD Figure-of-Merit (FoM) value was optimized to 2.45 at the gamma-equivalent energy region of the thermal neutron by using the charge comparison method. However, this value was reduced to 1.37 at the lower gamma-equivalent energy region of more than 325 keV, and neutrons were difficult to distinguish from gamma rays. Hence, new algorithms should be studied to improve the PSD performance at low gamma-equivalent energy regions. Convolutional Neural Networks (CNNs) have excellent image recognition capabilities, and thus, neutron and gamma-ray waveforms can be discriminated by their characteristics through a known training set. In this study, neutron and gamma-ray waveforms were measured with a 137Cs source and moderated 252Cf source via an SiPM array-coupled CLYC detector and divided into two groups: training and PSD testing. The CNN training set comprised 137Cs characteristic gamma-ray waveforms and thermal neutron waveforms that were discriminated by the charge comparison method from the training group. A CNN with two convolution-pooling layers was designed to accomplish PSD with the test group. The PSD FoM value of the CNN method was calculated to be 37.20 at the gamma-equivalent energy region of more than 325 keV. This result was much higher than that of the charge comparison method, indicating that neutrons and gamma rays could be better distinguished with the CNN method, especially at low gamma-equivalent energy regions.

Effect of Using Different Data Libraries and Simulation Codes on the Calculation of Spectra and Operational Quantities for the D2O-252Cf Source at PTB

The neutron reference field produced by a heavy-water moderated 252Cf source is used at PTB for calibrating neutron-measuring devices. Knowledge of the precise neutron spectrum is very important for the investigation of operational radiation protection quantities. Recently, new fission spectrum data of 252Cf has been proposed based on the latest nuclear data library version. At PTB, earlier calculations for the D2O moderated 252Cf neutron source were carried out more than 20 years ago, thus updated and more detailed calculations are required. In this paper, a detailed simulation model of the PTB moderated 252Cf source assembly has been prepared and investigated using new spectral data and two different Monte Carlo transport codes MCNP6.1 and PHITS3.22, with ENDF/BVIII.0, ENDF/B-VII.1, ENDF/B-V, and ENDL85 evaluated nuclear data libraries. The results show that the evaluated nuclear data libraries influence the calculated operational quantities by (3-5) %. The dosimetric quantities calculated with the PHITS code and the ENDF/B-VII.1 data library agree well with the MCNP6 results.

Experimental study of boron-coated straws with a neutron source

Multiple types of high quality neutron detectors are proposed for the first phase of the China Spallation Neutron Source (CSNS), which will be commissioned in 2018. Considering the shortage of 3He supply, a detector module composed of 49 boron-coated straws (BCS) was developed by Proportional Technologies Inc. (PTI). Each straw has a length of 1000 mm and a diameter of 7.5 mm. Seven straws are tightly packed in a tube, and seven tubes are organized in a row to form a detector module. The charge division method is used for longitudinal positioning. A specific readout system was utilized to output the signal and simultaneously encode each straw. The performance of this detector module was studied using a moderated 252Cf source at the Institute of High Energy Physics (IHEP). The signal amplitude spectrum indicates its n-gamma discrimination capability. Despite the complex readout method, a longitudinal resolution of FWHM=6.1 ± 0.5 mm was obtained. The three-dimensional positioning ability qualifies this BCS detector module as a promising detector for small angle neutron scattering.

Large-Area, Low-Cost, High-Efficiency Neutron Detector for Vehicle-Mounted Operation

We have developed a large-area, low-cost, high-efficiency neutron detector for vehicle-mounted operation. The detector, which has overall dimensions 12.7 cm $\times 91.4$ cm $\times 102$ cm (5” $\times 36$ ” $\times 40$ ”), a sensitive area equal to $0.85~\text {m}^{{\mathbf {2}}}$ (1320 in2), and weight of 110 kg (242 lbs), employs an array of 90 boron-coated straw (BCS) detectors. PTI has also developed electronics to minimize cost and space while providing low-noise signal conditioning for both neutron and gamma detection channels, as well as low energy Bluetooth communication with handheld devices. Extremely low power consumption allows continuous use for 225 hours (~10 days) using three AAA lithium-ion rechargeable batteries. We present radiological, mechanical, and environmental tests, collected from four full-scale prototypes. Outdoor neutron-counting tests with a moderated 252Cf source 2 m away from the center of the detector face showed an average detection rate of 5.5 cps/ng with a standard deviation of 0.09 cps/ng over the four individual detector measurements. Measurements showed a gamma rejection ratio of $1.0\times 10^{{\mathbf {-8}}}$ , and gamma absolute rejection ratio (GARRn) of 0.93. The prototypes were also operated successfully onboard a moving vehicle for high-speed tests and a long-range 1433-mile, two-day road trip from Houston, TX, USA, to Laurel, MD, USA. Using auxiliary DARPA SIGMA equipment, the GPS, timestamp, gamma and neutron data were transmitted over the cellular network with 10 Hz resolution to a server and real-time tracking website. Mechanical impact and electrostatic discharge testing produced no spurious counts in either the neutron or gamma channels. Ambient environmental temperature testing showed less than ± 1% response variation over the range from −30°C to +55°C.

Study of the thermal neutron detector ZnS(Ag)/LiF response using digital pulse processing

The neutron and gamma responses of a commercial 6Li loaded ZnS(Ag) detector (EJ-420) have been studied using a moderated 252Cf source and a set of pure gamma ray sources (137Cs and 241Am). Digital pulse processing was used to implement the pulse shape discrimination technique in order to well-separate neutrons from gamma-induced events. It was found that in case of a mixed neutron/gamma radiation field (252Cf) pulse-height discrimination technique works quite good for the measurement of the neutron field, but if one desires to reach a gamma-to-neutron sensitivity ratio lower than 0.7% pulse shape discrimination technique must be implemented. The decay time of the scintillation components of ZnS(Ag)/6LiF were determined by following the so called ``peeling off method''. The results were compared with some reported values. In order to better understand the shape of the neutron-induced pulses produced by the EJ-420 detector, a measurement with an EJ-440 detector of an alpha source (210Po) was performed. The response of both detectors must be very similar because these are based in ZnS(Ag) scintillator. It seems that for neutron detection with the EJ-420, the slow component of the scintillation yield produced by the tritium is higher than the one produced by the alpha particle. Finally, the neutron efficiency and some experimental features of the EJ-420 detector were determined and reproduced respectively by Monte Carlo simulations (GEANT4 9.6).

Graphite moderated 252Cf source

The Thorium molten-salt reactor is an attractive and affordable nuclear power option for developing countries with insufficient infrastructure and limited technological capability. In the aim of personnel training and experience gathering at the Universidad Simon Bolivar there is in progress a project of developing a subcritical thorium liquid-fuel reactor. The neutron source to run this subcritical reactor is a 252Cf source and the reactor will use high-purity graphite as moderator. Using the MCNP5 code the neutron spectra of the 252Cf in the center of the graphite moderator has been estimated along the channel where the liquid thorium salt will be inserted; also the ambient dose equivalent due to the source has been determined around the moderator.

Development of a neutron standard field using a heavy-water moderated 252Cf source at NMIJ-AIST

Development of a neutron standard field using a heavy-water moderated 252Cf source at NMIJ-AIST

Mixed Lutetium Iodide Compounds

In this paper we present initial results on mixed LuI3-YI3-GdI3 scintillators for gamma and neutron detection. The scintillation properties were investigated and compared to the results obtained for the binary compositions. Small samples, few millimeters in size, were tested. Under X-ray excitation each crystal exhibits bright green emission in the 425 to 750 nm spectral range. The exact peak position depends slightly on the material composition. The emission is due to d-f transitions on the Ce3+ ion. The scintillation under gamma excitation is fast and decays with about 30 ns time constant. The rise time varies slightly and it can be as fast as 0.5 ns. The light output of the investigated samples varied from 115,000plusmn13,000 ph/MeV for LuI3 to 68,000plusmn8,000 ph/MeV for LuI3-YI3 composition. The energy resolution at 662 keV was difficult to estimate since the samples lacked in crystal quality. LuI3-GdI3 composition was also tested for the detection of thermal neutrons from a moderated 252Cf source. The neutron peak appears at 81 keV when compared to 60 keV peak in 241Am spectrum recorded under the same conditions. The light output per a single detected neutron is about ~6,350 photons.

An Inter-comparison of the Neutron Calibration Fields by D2O Moderated 252Cf Source at JAEA and KAERI

An inter-comparison of the neutron calibration field of the D2O moderated 252Cf source was carried out at JAEA and KAERI. Measuring the neutron spectra at KAERI by means of the method employed at JAEA, we attempted to evaluate the basic parameters for the D2O moderated 252Cf source and validate our method for determination of these parameters. We obtained the differences in the spectrum at each facility and theoretically clarified the effect of the geometries of the source assembly and the dimensions of the irradiation room on the neutron standard field.

Calibration of neutron albedo dosemeters.

It is shown that by calibrating neutron albedo dosemeters under the proper conditions, two complicating effects will essentially cancel out, allowing accurate calibrations with no need for explicit corrections. The 'proper conditions' are: a large room (> or = 8 m on a side). use of a D2O moderated 252Cf source, and a source-to-phantom calibration distance of approximately 70 cm.

Use of gamma-ray spectroscopy for direct detection of thermal neutrons

A neutron detector based on detection of the prompt gamma-radiation emitted as a result of neutron interaction with a suitable neutron absorber is described. Boron-10 loaded polyethylene is used as the neutron converter. Neutrons are thermalized in the polyethylene and are subsequently absorbed by 10B. The result of the neutron absorption is production of 7Li, in the excited state (7Li*) and an alpha-particle. This occurs for 94% of the absorbed neutrons. The 7Li* (T1/2 = 10-13 s) emits a 477.6 keV gamma-ray ("prompt" gamma-ray) after it is formed and may be detected in a gamma-ray spectroscopy system. Since the "prompt" gamma-ray peak is Doppler broadened – the recoiled 7Li* emits the gamma-ray on "fly" – it is easily distinguishable from other non-capture gamma-ray peaks. The neutron converter, 10B loaded into a polyethylene end cap placed on the standard gamma-ray spectroscopy detector, adds the neutron detection capability without impeding the system's gamma-ray spectroscopy characteristics. Results for detection of neutrons from a moderated 252Cf source are presented.

Experimental optimisation of a moderated 252Cf source for land mine detection

A moderated 252Cf source to be used for the detection of buried land mines has been experimentally characterised. The moderating structure was obtained by using an inner sphere of Pb embedded in a high-density polyethylene (HDPe) brick assembly. The number of capture events in a Cd sample placed outside the structure was used, by varying the dimensions of the brick assembly, to determine the best moderator geometry. Furthermore, the detection of land mines by neutron capture reaction on nitrogen nuclei contained in the explosive was simulated by in field measurements placing the Cd sample at different depths in the soil. The obtained results are compared with predictions from Monte Carlo calculations.

Properties of Threshold Detectors for Testing Neutron Dosimetry Calibrations

Abstract A neutron detector combination based on solid state nuclear track detectors, which makes use of uranium oxide and LiF converters together with cadmium plastic absorbers was calibrated in a known neutron field of a heavy water moderated 252Cf source. From the derived equations describing the spectrometer characteristics it is possible to measure the average 235U neutron fission cross section s5 in the energy interval from 10 to 104 eV where there are more than 50 resonances. s5 was obtained from fission track densities in two Makrofol detectors, a bare one and one with cadmium shielding. The experimental s5 values of 5.7 and 20.5 b agree with the calculated values of 1.3 to 22 b obtained from a nuclear data code. The reported average cross section values cannot be considered as new nuclear data, but the utility is shown of using them for data analyses.

Calibration of various types of neutron dosimeters in a heavy water moderated and monoenergetic neutron fields-I. theoretical

Computer calculations have been performed to determine calibration factors of various types of neutron dosimeters in a heavy water moderated 252Cf and monoenergetic neutron fields, and the mediums being free air and a water phantom. In the case of heavy water moderated 252Cf source, the neutron spectra from the source was determined with MORSE-CG code. The backscattered neutron spectra on the surface of the phantom was calculated with MNCP code and then the contribution of these backscattered neutrons to the dose equivalent was estimated.

Development of 3He solid state detectors epithermal neutron spectrometer

A neutron spectrometer was developed for measuring the neutron energy in the range from 1–300 keV. Its fabrication was based on the 3He(n,p) 3H reaction, with two Si solid state detectors used for measuring the energy of the two charged particles. The sum of the energies of the two particles is the neutron energy E n plus the reaction Q-value (764 keV). The FWHM detector resolution is ≈3%. The detector efficiency, ≈3% for thermal neutrons, is proportional to 1 E (i.e., to the detection reaction cross section). Cooling improves the resolution. To avoid obscuring of the low keV energy range by thermal neutrons, the detector was covered with a 1 mm layer of Cd. Background due to neutron interaction with the Si detector, and with the detector construction materials and gamma rays, was removed by repeating the foreground measurement with no 3He in the detector. Neutron spectra were measured both with a moderated 252Cf source and with an epithermal beam at the Brookhaven Medical Research Reactor (BMRR). Results were in good agreement with Monte Carlo calculations using the MCNP code.

Intercomparison of Dose Equivalent Meters Based on Microdosimetric Techniques: Detailed Analysis and Conclusions

An intercomparison of area monitors based on microdosimetric techniques was jointly organised by EURADOS and Physikalisch-Technische Bundesanstalt (PTB), Braunschweig. Instruments based on tissue-equivalent proportional counters (TEPCs) from seven European Institutes were irradiated at PTB in neutron reference fields (monoenergetic neutrons in the range from thermal neutrons to 14.8 MeV neutrons, the radiation field of a D2O moderated 252Cf source) and at a 60Co gamma ray reference source. The results of the intercomparison have been presented in other reports. In this paper they are analysed in detail. The analysis takes advantage of the fact that several of the participating systems provided measured microdosimetric spectra as part of the results output. The variation of the dose equivalent response with neutron energy and the differences in response between different systems is an important aspect of the analysis. The effects of calibration and counter geometry on the quantities measured as well as on the diagnostic information in terms of Q values and neutron-photon dose discrimination are discussed. Of particular practical relevance is the evaluation of statistical uncertainties and thus sensitivities of TEPCs. Two methods of measurement, pulse height analysis and the variance technique are compared. The results show that TEPC based area monitors can be made to have a dose equivalent response which varies less with neutron energy than existing radiation protection dosemeters. The analysis shows that further improvement is possible by optimising operation, calibration and evaluation procedures.

Intercomparison of Dose Equivalent Meters Based on Microdosimetric Techniques: Detailed Analysis and Conclusions

An intercomparison of area monitors based on microdosimetric techniques was jointly organised by EURADOS and Physikalisch-Technische Bundesanstalt (PTB), Braunschweig. Instruments based on tissue-equivalent proportional counters (TEPCs) from seven European Institutes were irradiated at PTB in neutron reference fields (monoenergetic neutrons in the range from thermal neutrons to 14.8 MeV neutrons, the radiation field of a D2O moderated 252Cf source) and at a 60Co gamma ray reference source. The results of the intercomparison have been presented in other reports. In this paper they are analysed in detail. The analysis takes advantage of the fact that several of the participating systems provided measured microdosimetric spectra as part of the results output. The variation of the dose equivalent response with neutron energy and the differences in response between different systems is an important aspect of the analysis. The effects of calibration and counter geometry on the quantities measured as well as on the diagnostic information in terms of Q values and neutron-photon dose discrimination are discussed. Of particular practical relevance is the evaluation of statistical uncertainties and thus sensitivities of TEPCs. Two methods of measurement, pulse height analysis and the variance technique are compared. The results show that TEPC based area monitors can be made to have a dose equivalent response which varies less with neutron energy than existing radiation protection dosemeters. The analysis shows that further improvement is possible by optimising operation, calibration and evaluation procedures.