Sensitivity Study of PADC Track Detector with External Radiators
The aim of this study was to investigate the sensitivity and reliability of PADC as a neutron dosemeter suitable for large scale routine personnel monitor. A personal neutron dosemeter is being developed in our laboratory by using PADC as a detector and hydrogenated materials as proton converters for fast neutron detection. To increase the PADC response and sensitivity to thermal neutrons, boron converter was added to the dosemeter. Several dosemeters with these configurations were mounted on a water-filled phantom and exposed normal to: 241Am-Be, 241Am-Be softened with a 20 cm radius of polyethylene sphere moderator, 252Cf neutron spectra. The irradiated detectors were electrochemically etched and evaluated in order to determine their dose equivalent response in terms of H (10, α). The results obtained were compared to those obtained from Monte Carlo simulations using the MCNPX code in order to improve, the sensitivity of the PADC response. The detector configuration can be adapted to neutron spectra in different practical situations. It would be able to provide dose equivalent response in a large range of energies.
Received: July 19, 2010;
Accepted: September 09, 2010;
Published: October 14, 2010
Around the reactors and particle accelerators which used in research, industry
and radiation therapy, there is possibility to detect thermal neutrons as secondary
radiation, which is produced in the interactions with the structural materials
of the accelerator and/ or with targets, beam dumps and collimators influence
the shielding design. For low and intermediate energy accelerators, neutrons
are the particles that contribute mainly to these secondary radiation fields
(Silari, 2001). Thermal neutrons can be detected via
(n,α) reactions in 10B, because low energy α particles
≤2 MeV could be recorded. The Poly-Allyl Diglycol Carbonate (PADC) detector
is of particular interest for the development of a fast neutron dosemeter (Griffith
et al., 1981). Neutron elastic interactions with CR-39 plastic leave
latent recoil charged particle tracks which can be brought out by chemical (EC)
or electrochemical etching (ECE). The track density and the geometrical parameters
of tracks depend on the track formation process, which is strongly determined
by the type of particles and the etching conditions applied to develop the latent
tracks (Tommasino and Harrison, 2004). Above a threshold
of ~7.9 MeV, inelastic alpha particle breakup of 12C is also possible
with the energy lost by the neutron partitioned among the three alpha particles
(Fleischer et al., 1965). Fast neutrons interact
with the constituents of the CR-39 detector and produce H, C and O recoils as
well as (neutron (n), alpha (α)) reactions. These neutron- induced charged
particles contribute towards the response of CR-39 detectors (Khan
et al., 2000). Study of its application to neutron dosimetry has
shown that this material can be used in personnel neutron dosimetry. Using combination
of chemical pre-etching and electrochemical etching may lead to detect low energy
neutrons (Azimi-Garakani et al., 1987). So there
is the need to apply especial electrochemical etching condition. ECE in combination
with automatic image processing systems is widely used in neutron dosimetry
(Tommasino et al., 1984). The size and characteristics
of the tracks depend upon the charged particle mass, energy and direction. Energy
dependence has been reported previously on the distribution of track sizes for
monoenergetic neutrons (Hankins and Westermark, 1987)
and for broad spectrum neutrons (Jakes et al., 1997).
Phillips et al. (2006) have investigated the
dependence of the track density with energy, fluence and with direction. PADC
in contact with boron and polyethylene have been studied to provide a thermal
sensitive neutron detector. The aim of this research is to study the high sensitivity
and reliability of PADC as a neutron dosemeter suitable for large scale routine
personnel monitor. This study has been performed with new configuration of the
thermal converter, which were simulated by MCNPX code (MCNPX,
2002) and validated with a series of irradiations to realistic neutron fields.
MATERIALS AND METHODS
The study was conducted from Sep - 2008 to May - 2010 in our neutron dosimetry laboratory.
Dosemeter arrangement: It is composed, following the incidence direction of the neutron beam, of a 3 mm thick layer of polyethylene acting as fast neutron converter via elastic scattering with hydrogen, on a 1 mm thick layer of PADC used as a detector and on a 5 mm thick methacrylate holder. To increase the sensitivity of this configuration to thermal neutrons, 250 μm of boron layer was added next to the polyethylene layer. The cadmium absorber with 250 μm thick was added rear surface of the PADC. The whole configuration can be seen in Fig. 1.
Neutron irradiation: Thirty irradiation cards of dosemeter approximately 1.8x1.8 cm2 were irradiated for 241Am-Be, 241Am-Be softened with a 20 cm radius of polyethylene sphere moderator and 252Cf neutron spectra. Irradiations were performed with the dosemeters mounted on water-filled phantoms of 15x30x30 cm3 at normal neutron incidence.
Processing and reading: The PADC plates were etched using an electrochemical
etching in optimized condition as specified in Table 1. The
etched PADC detectors were digitized using a high resolution scanner (2000 dpi)
with transparency adaptor (UMAX scanner power look III), which is connected
to a PC-based image analyzer and counting program. The program has been developed
in MATLAB 6.5.1 environment (Taheri, 2005). After implementing
the image processing, the tracks were automatically counted inside the Region
of Interest (ROI) and the track density was calculated. The ROI includes all
the tracks inside the scanning area (3 cm2) whose major axis ranges
greater than 20 μm. The background track density was measured over 18 samples
and the average value obtained was 40±5 cm-2.
|| Arrangement of the dosemeter configuration
|| The four steps of the electrochemical etching applied
Monte carlo simulations: The response function of detector with each converter was simulated with the MCNPX code. The code has been used to calculate the number of protons crossing the front detector surface and classify them according to their energy and angle of incidence. For this purpose, the Particle Track Output Card (PTRAC) was used for calculating recoils for the whole history. The PTRAC generates an output file of user-filtered particle events. This option of MCNPX code registers the whole history, such as interaction type, energy, direction and position of particles. The simulated source was a broad parallel beam of monoenergetic neutrons normal to detector surface with 33 energies ranging from 0.025 eV to 30 MeV. For all calculations, the code was run for 107 histories.
RESULTS AND DISCUSSION
To simulate the response of dosemeter configuration, the MCNPX code (MCNPX,
2002) has been used to calculate the number of protons, carbons and oxygen
crossing the front detector surface and produce in the PADC. To simulate the
real condition, recoils classify according to their energy and angle of incidence.
Improvement on thermal neutron up-scatter in polyethylene is performed by using
the S (α, β) model treatment (MCNPX, 2002).
To improve the sensitivity of detector, boron layer has been used in contact
with front surface of the PADC as shown in Fig. 1. The detector
response is obtained by considering protons, carbon, oxygen recoils and a particles
with incidence angles smaller than the critical angle of 45° and with energies
within the PADC energy windows response 0.1-2.4 MeV (Fernandez
et al., 1996). The simulated response function of the PADC is compared
with the result which is presented in the IAEA technical reports series No.
318 (IAEA, 1990). The shape of the responses in Fig.
2 shows a good agreement between simulation and the results have been processed
electrochemically (IAEA, 1990). The experimental and calculated
average responses for all neutron sources which have been used in this study
also show good agreement. The response function curves of the PADC in contact
with polyethylene and boron convertor are plotted in Fig. 3
and 4, respectively. For each shape, the response is simulated
with MCNPX shows that the sensitivity of the detector to epithermal and thermal
neutron spectra is increased. These figures shows that the sensitivity of the
detector for low energies of the incident neutron is increased in compare to
results presented (Phillips et al., 2006; IAEA,
1990). The Y axis of the Fig. 2 to 4
are expressed in arbitrary units, but the magnitude of the units is the same
for all graphs. The integrated PADC detector with different converter which
has been presented in this work shows good improvement of the PADC response
to thermal neutron. The code can be used to define the most adequate dosemeter
configuration adapted to neutron spectra in practical situations encountered
at protection level. In order to compare the simulated result with experiment,
the energy used was the same as that for the experimental irradiations.
The experimental neutron response in terms of Hp (10, α) has
been evaluated from the net measured track densities (measured track density
minus average background) taking into account the corresponding reference values
of each source corrected for the effect of source-to-detector distance.
|| Comparison of the response function curve of the PADC detector
|| Simulated response function curve of the PADC + Polyethylene
|| Simulated response function curve of the PADC + Boron converter
||Experimental and simulated dose equivalent response and normalized
per unit personal equivalent dose
experimental and simulated dose equivalent response and normalized per unit
personal equivalent dose at normal incidence and for the calibration sources
are shown in Table 2, which also displays the uncertainties
corresponding to one standard deviation. A very good agreement can be seen between the simulated and the experimental
results of the dosemeter configuration for all sources. The response value is
the lowest for 241Am-Be and increases as energy of neutrons moderated
and for 252Cf was reached to the highest value. This fact can be
attributed to a detection efficiency increase when the energy of the source
decreases. In order to use this dosemeter in routine measurements, a calibration
factor, (2.7±0.9) μSv.cm2, for all sources and neutron
incidences has been calculated as the reciprocal of the mean dose equivalent
response to 0o neutron incidence. The percent deviations between
the values obtained using the calibration factor and real doses in all the cases
range from +49% and -11%, which fulfill the International Atomic Energy Agency
(IAEA) requirement. The corresponding limit (-33%) was recommended by the IAEA
(1999). The results of detector configuration which is presented in this
study, is comparable with the data presented by Garcia et
al. (2005). The minimum detectable dose equivalent (MDDE) is defined
with a level of confidence of 97.5% according to Harvey
et al. (1998). A mean MDDE value of 62±25 μSv has been
found for all sources and neutron incidence with 0° angle. This value is
smaller than the limit of 80 μSv recommended by the International Commission
on Radiological Protection (Annals of the ICRP 60, 1990).
Monte Carlo simulations showed that the shape of the response function of the PADC is very similar to that is presented in IAEA Safety Guides, No. RS-G-1.3. The small discrepancies can be ascribed to actual critical angle and geometrical effects. From the results of this work, it can be stated that the MCNPX computer code reproduces the experimental results with good agreement. The code can also be used to define the most adequate dosemeter configuration for thermal and intermediate neutrons if the cross sections of the appropriate nuclear reactions are introduced in the code. The detector configuration can be adapted to neutron spectra in different practical situations. It would be able to provide dose equivalent response in a large range of energies, as well as for intermediate neutrons.
Deviations of the dose equivalent evaluated using the mean calibration factor
obtained in this work from its true value range from -11 to +42%, which fulfill
the IAEA requirement. The minimum detectable dose equivalent of the dosemeter
which is presented in this work has been found to be 62±25 μSv,
which is smaller than the 80 μSv limit recommended by the ICRP 60. This
good value is probably due to having the PADC with good condition, converter
and electrochemical etching method, which is known as an optimized technique.
The authors are very grateful to the staff at the neutron laboratory of the national radiation protection dosimetry, in particular to Mr. A. Kazemi Movahed, Mr. M.R. Sadegh khani and Mr. A. Kamali for their valuable help in arrangement of the experiments.
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