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Optimization of moderator assembly for neutron flux measurement: Experimental and theoretical approaches

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Abdul Waheed
Abdul Waheed
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Nawab Ali at Pakistani Institute of Nuclear Science and Technology
Nawab Ali
Muzahir Ali Baloch at National University of Technology, Islamabad
Muzahir Ali Baloch
  • National University of Technology, Islamabad
Aziz ahmed Qureshi at COMSATS University Islamabad
Aziz ahmed Qureshi
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Abstract

A moderator of paraffin wax assembly has been demonstrated where its thickness can be optimized to thermalize fast neutrons. The assembly is used for measuring fast neutron flux of a neutron probe at different neutron energies, using BF3 (Φ1″ and 2″) and 3He(Φ0.5″) neutron detectors. The paraffin wax thickness was optimized at 6 cm for the neutron probe which contains an Am–Be neutron source. The experimental data are compared with Monte Carlo simulation results using MCNP5 version 1.4. Neutron flux comparison and neutron activation techniques are used for measuring neutron flux of the neutron probe to validate the optimum paraffin moderator thickness in the assembly. The neutron fluxes are measured at (1.17 ± 0.09) × 105 and (1.19 ± 0.1) × 105 n/s, being in agreement with the simulated values. The moderator assembly can easily be utilized for essential requirements of neutron flux measurements.
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Optimization of moderator assembly for neutron flux
measurement: experimental and theoretical approaches
Abdul Waheed
1
•Nawab Ali
2
•Muzahir A. Baloch
3
•Aziz A. Qureshi
1
•
Eid A. Munem
4
•Muhammad Usman Rajput
2
•Tauseef Jamal
5
•Wazir Muhammad
6
Received: 24 June 2016 / Revised: 2 September 2016 / Accepted: 24 September 2016 / Published online: 28 March 2017
Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Chinese Nuclear Society, Science Press China and Springer
Science+Business Media Singapore 2017
Abstract A moderator of paraffin wax assembly has been
demonstrated where its thickness can be optimized to
thermalize fast neutrons. The assembly is used for mea-
suring fast neutron flux of a neutron probe at different
neutron energies, using BF
3
(U100 and 200) and
3
He(U0.500)
neutron detectors. The paraffin wax thickness was opti-
mized at 6 cm for the neutron probe which contains an
Am–Be neutron source. The experimental data are com-
pared with Monte Carlo simulation results using MCNP5
version 1.4. Neutron flux comparison and neutron activa-
tion techniques are used for measuring neutron flux of the
neutron probe to validate the optimum paraffin moderator
thickness in the assembly. The neutron fluxes are measured
at (1.17 ±0.09) 910
5
and (1.19 ±0.1) 910
5
n/s, being
in agreement with the simulated values. The moderator
assembly can easily be utilized for essential requirements
of neutron flux measurements.
Keywords Am–Be and
252
Cf neutron sources
BF
3
&
3
He detectors Paraffin wax Neutron flux
Monte Carlo simulation
1 Introduction
The Am–Be and
252
Cf neutron sources are widely used
in neutron dosimetry laboratories. The important features
of Am–Be source are its long half-life and emission of fast
neutrons with rather wide energy spectra, while
252
Cf
yields a highly concentrated and reliable neutron spectrum
from a small assembly. The last feature of both sources is
helpful in covering the energy domain of interest for many
applications in ambient and personal dosimetry. Never-
theless, the neutron energy spectrum depends on materials
and dimension of the capsule and on amount and physio-
chemical properties of active material, thus affecting rele-
vant quantities such as spectrum-averaged fluence-to-dose
equivalent conversion coefficient [1–6]. It is important to
know the neutron flux accurately so as to measure different
quantities related to a neutron source [7].
Neutron detection has a key role in the development of
nuclear technology and its applications [8,9]. High-sensi-
tive neutron detectors are required. However, most of the
neutron detectors are also sensitive to gamma rays.
Therefore, an effective technique is required to eliminate
the gamma background for accurate neutron
&Wazir Muhammad
wazirm@hotmail.com
1
Radiation Physics Laboratory, Physics Department,
COMSATS Institute of Information Technology (CIIT),
Islamabad, Pakistan
2
Physics Division, Pakistan Institute of Nuclear Science and
Technology (PINSTECH), Islamabad 45650, Pakistan
3
Department of Physics, College of Science, Majmaah
University, Al-Zulfi, Saudi Arabia
4
School of Physics, University Science Malaysia,
11800 Gelugor, Penang, Malaysia
5
DCIS, Pakistan Institute Engineering and Applied Sciences
(PIEAS), Islamabad 45650, Pakistan
6
Health Physics Division (HPD), Pakistan Institute of Nuclear
Science and Technology (PINSTECH), Islamabad 45650,
Pakistan
123
NUCL SCI TECH (2017) 28:61
DOI 10.1007/s41365-017-0213-z
measurements. Generally, methods used to recognize
neutron interactions within a detector rely on second-order
effects. Two neutron interactions used for a variety of
thermal neutron detectors are the
10
B(n,a)
7
Li [9,10] and
6
Li(n,a)
3
H reactions [11,12]. The aparticles can be easily
detected.
Low atomic number materials, such as high-density
polyethylene (HDPE), which has a high concentration of
hydrogen, tend to have relatively high elastic scattering
cross sections for fast neutrons, and often (n, p) reactions
from fast neutrons interacting in hydrogen-filled materials
are manipulated for fast neutron detection. For detection of
thermal or low-energy neutrons, polymers-based SSNTDs,
i.e. solid-state nuclear track detectors (such as CR-39), can
be used only in combination with a neutron reactive film
which converts neutrons into other detectable radiations,
recoil nuclei [13,14]. Use of a BF
3
or
3
He neutron detector,
one can measure the optimum thickness of the moderator
accurately, which is essential to avoid excessive thermal
neutron absorption in the moderator. The neutron activa-
tion analysis can determine accurately the major and trace
elements in different samples [10–12,15]. In practice,
aromatic hydrocarbons are considered as the most radia-
tion-resistant hydrogenous substances and have properties
to moderate slow neutrons more effectively [16].
A paraffin wax moderator assembly has been developed
to find out optimum thickness of paraffin wax. It can be
utilized for fast neutrons from various neutron-producing
reactions on an accelerator or for measurement of neutron
yield from a Mather-type plasma focus device [17–20]. The
assembly has been utilized to find out neutron flux of a
neutron probe which contains Am–Be as a neutron source
[21]. Neutron flux comparison and neutron activation are
used to the measure neutron flux [22]. A 50-mCi Am–Be
neutron probe is used as a neutron source for the mea-
surement of moisture in soils [21].
The moderating materials should be of large scattering
cross section, small absorption cross section and large
energy loss per collision [23], and paraffin wax is just a
desired moderator. Its optimum thickness was calculated
for 3 MeV neutrons [22] and was experimentally deter-
mined at 7 and 2.5–4.5 cm for Am–Be and
252
Cf neutron
sources, respectively [24,25]. The results were based on
the maximum gamma ray yield due to neutron activation.
The optimum thickness of HDPE moderator was measured
at 4–6 cm for 2.8 MeV neutrons. For theoretical calcula-
tions, different codes such as MNCP, MNCPX and SRIM
are used [26,27]. The present paper is designed to deter-
mine the optimum moderator thickness that is required to
be wrapped around a BF
3
detector used in monitoring fis-
sile material. The experimental results are compared with
MC simulations for Am–Be neutron source using MCNP5
code version 1.40 [28].
2 Materials and methods
2.1 Theoretical calculations
The work of Goshal [22] is referred for theoretical cal-
culations. The slowing down process involves elastic
scattering by light nuclei like hydrogen and carbon in
paraffin wax (q=0.93 g/cm
3
)[5]. The average logarith-
mic decrease in neutron energy per collision is calculated.
The average number of collisions needed to thermalize fast
neutrons is also calculated. The transport, thermal neutron
mean free paths are calculated for paraffin wax, and finally,
the optimum moderator thickness is calculated at 6 cm for
3 MeV neutrons.
2.2 Experimental approach
The paraffin wax moderator assembly consists of two
parts as shown in Fig. 1. The first part consists of a
removable paraffin wax discs, neutron detector and a
neutron source (i.e. Am–Be/
252
Cf neutron sources). The
second part is a mounting frame with two parallel spring-
loaded U12 mm steel rods, each fixed at one end with the
frame. The other two ends are fixed to a movable plastic
ring around a 1.4-cm-thick paraffin wax disc. The purpose
of the plastic ring is to give strength to the paraffin wax
discs. A number of paraffin wax discs are used. Sized at
U27 cm 90.6–1.4 cm, each disc is mounted in a 6-mm-
thick plastic ring. The paraffin wax discs can be assembled
easily on the mounting frame, between two semi-cylin-
drical fixed pipes with the assembly at the same axes, one
for the neutron detector and the other for the neutron
source. The two parallel rods act as a guide for the paraffin
wax discs and keep the discs in close contact with each
other. The U27 cm discs keep the axis same with respect to
the axes of the neutron detector and the neutron source.
Paraffin wax discs can be added or removed for measure-
ment of optimum thickness of paraffin wax, where the
Fig. 1 Sketch of the paraffin wax moderator assembly
61 Page 2 of 6 A. Waheed et al.
123
neutron detector counts is the maximum. Two BF
3
(U100
and 200) and one
3
He (U0.500) neutron detectors are used for
measuring the optimum thickness with Am–Be or
252
Cf
neutron sources. The associated electronics with detector
assembly consisted of pre-amplifier, amplifier, single-
channel analyser and counter/timer manufactured by the
ORTEC.
The background counts per second were measured
without the neutron sources, and they were subtracted from
the actual neutron count rates emitted from neutron sour-
ces. For different paraffin wax thicknesses, counts emitted
from either Am–Be or
252
Cf neutron sources were taken for
100 s by using the BF
3
and one
3
He neutron detectors. For
both the Am–Be and
252
Cf neutron sources, the paraffin
wax thicknesses were increased to find the maximum
counts, and on further increase in thickness, the counts
started decreasing. Here, the maximum counts represent
optimum paraffin wax thickness that is required to ther-
malize fast neutron (of course, the process is the trade-off
between absorption and moderation). To minimize scat-
tering from the walls, the heavy shielding was placed about
4 m away from the source. The scattering contribution at
the detector position was checked with and without the
shielding, with the source at its position. The difference
was negligible, and hence, an evidence of no scattering
contribution from shielding walls was seen. Furthermore,
BF
3
or
3
He detectors can detect only slow or thermal
neutrons and the gases have very low interaction cross
section for fast or scattering neutrons.
The moderator assembly (Fig. 1) was also used for the
measurement of neutron flux emitted from the Am–Be and
252
Cf neutron sources to check the validity of the optimum
thickness of the paraffin wax moderator, using the neutron
flux comparison method and the neutron activation analysis
method [29].
2.2.1 Neutron flux comparison method
The comparison of counting rates was carried out,
between the standard Am–Be neutron source of known flux
(F
1
) and that from the neutron probe of unknown flux (F
2
)
under conditions insensitive to the energy of neutrons. Let
C
1
be the average counts per sec for F
1
recorded by a BF
3
(U100) neutron detector in contact with the 6-cm-thick
paraffin wax moderator and C
2
be the average neutron
counts per sec, with their background subtracted, from the
neutron probe under the same condition. The neutron flux
F
2
from the neutron probe is given by F
2
=F
1
C
2
/C
1
.
2.2.2 Neutron activation analysis method
The paraffin wax moderator assembly was used to
measure the neutron flux of the neutron probe through
activation analysis technique by using silver foil, and the
6 cm optimum thickness of the paraffin wax moderator was
validated.
Thermal neutron activation analysis is the absolute
method to determine the neutron flux of a neutron probe.
The silver isotopes of
108
Ag (T
1/2
=0.39 min) and
110
Ag
(T
1/2
=24.6 s) are produced through nuclear reactions of
107
Ag(n, c)
108
Ag and
109
Ag(n, c)
110
Ag with thermal
neutron. After a few half-lives of
110
Ag, most of the
radioactivity in silver foil is due to
108
Ag [30]. A G.M
counter with Mylar end window is used for the neutron
activation analysis. The silver foil is placed at the Mylar
window. Both the neutron probe and the end window of
G.M detector are placed at the same axis in opposite and in
contact with the paraffin wax moderator (Fig. 2).
The optimum paraffin wax thickness of 6 cm was used
in the moderator assembly. At the termination of activa-
tion, the flux uof the neutron probe (in n/cm
2
/s) is given
by u=[A
0
w(1 -e
-kt
)/(rA
av
am)] e(4p/X)[22], where A
0
is the activity of
108
Ag at t=0, wis the atomic weight of
silver, kis the decay constant, tis the irradiation time,
r=37.6 barns (taken from the ENDF library [31]) is the
thermal neutron capture cross section of
107
Ag, A
av
is the
Avogadro’s number, ais the natural isotopic abundance of
107
Ag, mis the mass in grams of
107
Ag, eis the intrinsic
efficiency of the detector, and Xis the solid angle sub-
tended by the G.M detector at the neutron source (Am–Be)
of the probe.
The intrinsic efficiency of the G.M detector was deter-
mined from a
137
Cs c-ray source of known activity. The
silver foil was activated for 5 min. Although both
108
Ag
and
110
Ag are produced, most of the activity is due to
108
Ag
at the time of counting. The neutron probe was taken far
away from the moderator assembly as quickly as possible,
so the G.M detector background count was made almost
negligible at the time of counting. The activated silver foil
was counted immediately and was repeated several times
for better results. The background of the G.M detector was
measured in advance.
Fig. 2 The G.M detector with the end Mylar window, Ag foil, the
paraffin wax moderator and the neutron probe
Optimization of moderator assembly for neutron flux measurement: experimental and theoretical…Page 3 of 6 61
123
2.3 Monte Carlo simulation
MCNP is an advanced Monte Carlo code for simulation
of the neutron transport [28,32]. It has the ability to model
complex geometry of designed experiment and contains all
necessary cross-section data for neutron, photon and elec-
tron transport calculations [32]. For a neutron detection
process, the history of each starting neutron is followed
between collisions, and its energy deposition is recorded,
throughout its life to its death, until its energy is low
enough to be neglected. In the present study, the experi-
mental set-up was simulated in the School of Physics,
University Science Malaysia, Penang, Malaysia, using the
MCNP5 code version 1.40 [28,32]. The experimental set-
up as described in Fig. 1was modelled. According to this
model, the neutron source was located at one side of the
paraffin wax moderator, while the neutron detector was
simulated facing the source at the other side of the mod-
erator. The experimental set-up was simulated on wooden
table top. To calculate the detector neutron flux, F
2
tally
calculates the flux averaged over a surface [28] was used.
Surface segmentation was utilized to calculate the flux at a
surface segment facing the neutron source.
3 Results and discussion
The measured counts per 100 s for different wax
thicknesses by using the three neutron detectors for the
Am–Be and
252
Cf neutron sources plotted against wax
thickness are shown in Fig. 3. The maximum counts for
each neutron detector and source represent the optimum
paraffin wax thickness that is required to thermalize fast
neutrons. From Fig. 3, the optimum paraffin wax thick-
nesses were 6 and 5.5 cm for Am–Be and
252
Cf neutron
sources, respectively. The yield of thermal neutrons
increases with the thickness of paraffin wax from 1.6 to
6 cm, due to fact that the fast neutrons lose their energy
through elastic collision with hydrogen atoms of paraffin
wax. The yield of thermal neutrons at these thicknesses is
found to be maximum, as a large number of fast neutrons
lose their maximum energy through elastic collisions.
However, on further increase in thickness of the paraffin
wax from 6.5 to 12 cm, the yields of thermal neutron
decrease due to elastic and inelastic scattering reactions.
The neutrons lose their energy through the interactions
until they are captured by paraffin wax (shielding mate-
rial) [23]. Normally, the neutron capture cross section is
larger only for thermal neutron energies, that is why the
neutrons slowing down by scattering are important before
capture [23].
To validate the optimum thickness of paraffin wax, the
neutron flux was calculated. By the neutron flux compar-
ison method, the average neutron counts per second, C
1
,
was 141.3 ±0.66 with the known neutron flux of
F
1
=(8.3 ±0.66) 910
4
n/s. With C
2
=199.1 ±0.79 of
the neutron probe, the F
2
was (1.17 ±0.09) 910
5
n/s. On
the other hand, the neutron flux deduced by the activation
method was (1.19 ±0.1) 910
5
n/s. The two results are
virtually the same, considering the measurement errors.
The neutrons emerging out of paraffin wax are of both
thermal and epithermal in energy. The fluence of epither-
mal neutrons was estimated from Ref. [30]. The ratio of
epithermal to the total neutrons at a distance of 18.9 cm in
air from the neutron source is 5.56% where the total flu-
ence is maximum as calculated from Ref. [33]. Therefore,
this ratio is the same for the paraffin moderator at 6 cm in
thickness.
By using MCNP5, the energy spectra of Am–Be neutron
source at 3, 5 and 8 MeV were produced with the relative
probabilities of 1.000, 0.986 and 0.551, respectively. The
simulation results for the Am–Be source are shown in
Fig. 4, and the counts are maximized at 6 cm thickness of
the paraffin wax moderator, agreeing well with the exper-
imental results.
Fig. 3 The paraffin wax
thickness versus neutron counts,
with different neutron detectors
for Am–Be aand b
252
Cf
neutron probes
61 Page 4 of 6 A. Waheed et al.
123
4 Conclusion
The paraffin wax moderator assembly can be used easily
to thermalize fast neutrons of various energies produced
from neutron-producing reactions on an accelerator. Neu-
tron flux comparison and neutron activation techniques are
used for accurate determination of neutron flux of the
neutron probe to validate the optimum thickness of the
paraffin wax moderator in the assembly. The neutron flux
of (1.17 ±0.09) 910
5
n/s by the neutron flux comparison
method agrees well with the flux of (1.19 ±0.1) 910
5
n/s
by neutron activation analysis method. The techniques can
be utilized for the measurement of neutron yield from
Mather-type plasma focus device. The optimum thickness
of the paraffin wax moderator is 6 cm for Am–Be and
5.5 cm for
252
Cf neutron sources.
Acknowledgements The authors gratefully acknowledge services
provided by Mr. Usman Ali, RPL, CIIT, in the fabrication of
mechanical assembly and safe handling of neutron sources.
References
1. R. Bedogni, C. Domingo, N. Roberts et al., Investigation of the
neutron spectrum of americium-beryllium sources by Bonner
sphere spectrometry. Nucl. Instrum. Methods Phys. Res. Sect. A
763, 547–552 (2014). doi:10.1016/j.nima.2014.06.040
2. F.D. Becchetti, M. Febbraro, R. Torres-Isea et al.,
252
Cf fission-
neutron spectrum using a simplified time-of-flight setup: an
advanced teaching laboratory experiment. Am. J. Phys. 81,
112–119 (2013). doi:10.1119/1.4769032
3. S. Croft, K.A.Miller, Group representation of the prompt fission
neutron spectrum of
252
Cf. Esarda Bull., 112–114 (2011). https://
esarda.jrc.ec.europa.eu/images//Bulletin/Files/B_2011_046.pdf
4. J.W. Marsh, D.J. Thomas, M. Burke, High resolution measure-
ments of neutron energy spectra from Am-Be and Am-B neutron
sources. Nucl. Instrum. Methods Phys. Res. Sect. A 366, 340–348
(1995). doi:10.1016/0168-9002(95)00613-3
5. J.H. Wu, Y.J. Xu, S.L. Liu et al., Energy spectrum measurement
and dose rate estimation of natural neutrons in Tibet region. Nucl.
Sci. Tech. 26, 060202 (2015). doi:10.13538/j.1001-8042/nst.26.
060202
6. Z.G. Jiang, Y.G. Yuan, H.Y. Wang et al., Development of a
spherical tissue equivalent proportional counter for neutron
monitoring. Nucl. Sci. Tech. 26, 060403 (2015). doi:10.13538/j.
1001-8042/nst.26.060403
7. B. Wolle, R. BA
˜tzner, T. Baloui, Special absorber neutron
detector moderator assembly: a new detector system for flux
measurements of collimated 2.5 MeV neutrons. Rev. Sci.
Instrum. 70, 1194–1196 (1999). doi:10.1063/1.1149322
8. M.A. Rana, Research and applications of nuclear tracks: devel-
opments in Pakistan and global comparison. Nat. Sci. (2012).
doi:10.4236/ns.2012.431124
9. G.F. Knoll, Radiation detection and measurement (Wiley, New
York, 2000)
10. M.U. Rajput, N.L. Maidana, V.R. Vanin et al., Measurement of
thermal neutron cross section and resonance integral for the
165Ho(n, c) 166Ho reaction. Radiochim. Acta Int. J. Chem. Asp.
Nucl. Sci. Technol. 97, 63–69 (2009). doi:10.1524/ract.2009.1575
11. M. Akram, A. Iqbal, S.N. Husaini et al., Determination of boron
contents in water samples collected from the Neelum valley,
Azad Kashmir, Pakistan. Biol. Trace Elem Res. 139, 287–295
(2011). doi:10.1007/s12011-010-8665-6
12. K. Haygarth, D. Janik, I. Janik et al., Neutron and b/cradiolysis
of water up to supercritical conditions. 1. b/cyields for H
2
,H
atom, and hydrated electron. J. Phys. Chem. A 114, 5034 (2010).
doi:10.1021/jp101790p
13. M. Barbagallo, L. Cosentino, V. Forcina et al., Thermal neutron
detection using a silicon pad detector and
6
LiF removable con-
verters. Rev. Sci. Instrum. 84, 033503 (2013). doi:10.1063/1.
4794768
14. A.A. Naqvi, M.S. Abdelmonem, G. Al-Misned et al., New
source-moderator geometry to improve performance of
252
Cf and
241
Am-Be source-based PGNAA setups. Nucl. Instrum. Methods
Phys. Res. Sect. A 562, 358–364 (2006). doi:10.1016/j.nima.
2006.02.004
15. S. Ahmad, S.S. Hussain, M. Sadiq et al., Enhanced and repro-
ducible neutron emission from a plasma focus with pre-ionization
induced by depleted uranium (U238). Plasma Phys. Control.
Fusion 48, 745 (2006). doi:10.1088/0741-3335/48/6/003
16. V. Ananiev, A. Belyakov, M. Bulavin, et al., The world’s first
pelletized cold neutron moderator at a neutron scattering facility.
Nucl. Instrum. Methods Phys. Res. Sect. B 320, 70–74. (2014).
http://www1.jinr.ru/Preprints/2012/113(P13-2012-113).pdf
17. S.N. Ahmed, Physics and engineering of radiation detection
(Academic Press, New York, 2007)
18. M. Zakaullah, A. Waheed, S. Ahmad et al., Study of neutron
emission in a low-energy plasma focus with I
ˆ
2
-source-assisted
breakdown. Plasma Sources Sci. Technol. 12, 443 (2003). doi:10.
1088/0963-0252/12/3/320
19. M. Zakaullah, K. Alamgir, A. Rasool et al., Correlation of plasma
electron temperature with neutron emission in a low-energy
plasma focus. IEEE Trans. Plasma Sci. 29, 62–68 (2001). doi:10.
1109/27.912943
20. M. Zakaullah, I. Akhtar, A. Waheed et al., Comparative study of
ion, X-ray and neutron emission in a low energy plasma focus.
Plasma Sources Sci. Technol. 7, 206 (1998). doi:10.1088/0963-
0252/7/2/015
21. C. International, Neutron probe method using a 503DR
Hydroprobe, ICT, Editor, International Pty Ltd: Australia
22. S.N. Goshal, Nuclear Physics (S Chand Publisher, New Delhi,
1994)
23. M.H. Jasim, N.T. Abdulameer, Neutron capture cross section
measurements of paraffin wax. Int. J. Appl. Innov. Eng. Manag.
Fig. 4 The Monte Carlo simulation for the Am–Be neutron probe
versus the paraffin wax moderator thickness
Optimization of moderator assembly for neutron flux measurement: experimental and theoretical…Page 5 of 6 61
123
3, 112–114 (2014). http://ijaiem.org/volume3issue4/IJAIEM-
2014-04-20-034.pdf
24. V.M. Thakur, A. Jain, K. Biju, et al., Studies on optimization of
moderator thickness for BF
3
detectors used for monitoring of
fissile material. Indian J. Pure Appl. Phys. 50, 811–813 (2012).
http://nopr.niscair.res.in/handle/123456789/14910
25. A.A. Naqvi, M.M. Nagadi, M. Maslehuddin et al., Verification of
design calculations of a PGNAA setup using nuclear track
detectors. Radiat. Meas. 38, 37–41 (2004). doi:10.1016/S1350-
4487(03)00252-X
26. J.K. Zhao, J.L. Robertson, K.W. Herwig et al., Optimizing
moderator dimensions for neutron scattering at the spallation
neutron source. Rev. Sci. Instrum. 84, 125104 (2013). doi:10.
1063/1.4841875
27. J.F. Ziegler, M.D. Ziegler, J.P. Biersack, SRIM–The stopping and
range of ions in matter (2010). Nucl. Instrum. Methods Phys. Res.
Sect. B 268, 1818–1823 (2010). doi:10.1016/j.nimb.2010.02.091
28. R. A. Forster T.N.K.G., MCNP—A General Monte Carlo
N-Particle Transport Code, Version 5. 2003
29. M. Borsaru, P.L. Eisler, Neutron activation analysis (slow neu-
trons), in United States Patent, 1982
30. J.R. Parrington, D.K. Harold, L.B. Susan et al., Chart of
radionuclides—Knoll Atomic Power Laboratory (Naval Reactors,
US Department of Energy, Washington, DC, 1996)
31. M.B. Chadwick, M. Herman, P. Obloz
ˇinsky
´et al., ENDF/B-VII.
1 nuclear data for science and technology: cross sections,
covariances, fission product yields and decay data. Nucl. Data
Sheets 112, 2887–2996 (2011). doi:10.1016/j.nds.2011.11.002
32. X-5 Monte Carlo Team, MCNP—A General Monte Carlo
N-Particle Transport Code, Version 5 (2003)
33. M. Tohamy, M. Fayez-Hassan, M.M. Abdel-Khalek, et al., A
dual-hemisphere (spherical) irradiation facility for Am-Be neu-
tron field optimization (2009). http://www.iaea.org/inis/collec
tion/NCLCollectionStore/Public/42/009/42009903.pdf
61 Page 6 of 6 A. Waheed et al.
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... Since then, this method of neutron generation by irradiating light elements such as beryllium, lithium, and fluorine with an intense alpha source has been incorporated into radioisotopic neutron sources used in nuclear laboratories throughout the world. The mechanism by which the neutrons are produced proceeds via an (α, n) reaction which requires a source of high energy alpha particles [2,26,27], such as the 9 Be(α, n) 12 C reaction in beryllium. In particular, polonium-210 has been frequently used as the alpha source for such applications due to its high alpha emission energy of 5.4 MeV [30,31], as well as a high specific activity of 4 kCi/g, which stems from its relatively low atomic mass https://doi.org/10.32388/0HGLA6 ...
... Thus, a chain reaction can be initiated in a mixture of atoms of bismuth-209 and light elements, using an initial neutron source to transmute the bismuth atoms into that of polonium, which would themselves emit high-energy alpha particles that impact the light elements, causing further neutron production. Furthermore, 209 Bi and 9 Be are also able to cause neutron multiplication via a (n, 2n) reaction with fast neutrons [6], increasing the neutron population of the system. ...
... To be suitable for sustaining neutron multiplication, the bismuth salt should contain a light element with high neutron yields from (α, n) reactions [7], e.g. 27 Al, 19 F, 10 B, 9 Be, as well as a high hydrogen content for neutron moderation, as the 209 Bi neutron capture cross section for 210 Po formation is greatly increased at thermal neutron velocities. Hence, bismuth beryllium acetate was used due to its high beryllium content, which is ideal for alpha-to-neutron conversion, as well as its intrinsic neutron multiplication and moderation properties. ...
Neutronic Chain Reactions for Polonium-210 Production
Article
Full-text available
  • Oct 2023
The production of the industrially significant radionuclide polonium-210 from the neutron irradiation of bismuth metal and the subsequent beta decay of bismuth-210 is highly inefficient due to the small neutron capture cross section of bismuth-209. In this paper, we report a previously undescribed self-sustaining nuclear chain reaction involving self-propagating neutron multiplication in bismuth salts that allow for rapid and cost-effective production of polonium-210. The reaction proceeds in a cycle of three alternating elementary steps – the capture of neutrons by bismuth-209 and the subsequent formation of polonium-210, the emission of high-energy alpha particles by polonium-210, and the production of more neutrons from (α, n) and (n,2n) reactions on light element and bismuth-209 nuclei respectively. Furthermore, the high hydrogen density of the compound also confers it intrinsic neutron moderation properties, increasing the neutron capture cross section of bismuth-209 at thermal neutron energies. The chain reaction was proven to have successfully occurred by irradiating a sample of the bismuth salt with a 80 μCi neutron source and monitoring the activity levels of the reaction. It was found that the activity of the reaction increased exponentially after an initial stable period following a derived formula for polonium production trends for the reaction, thus validating the occurrence of the reaction. Furthermore, alpha spectroscopy confirmed that polonium-210 had been produced by characterising the 5.30 MeV alpha emission peak of the reaction in addition to using beta spectroscopy to identify the parent nuclide bismuth-210, further proving that the reaction was successful. Hence, this paper reports the successful initiation and characterisation of a novel nuclear chain reaction, and its potential applications offered by a method of rapidly producing large quantities of polonium-210.
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... Since then, this method of neutron generation by irradiating light elements such as beryllium, lithium, and uorine with an intense alpha source has been incorporated into radioisotopic neutron sources used in nuclear laboratories throughout the world. The mechanism by which the neutrons are produced proceeds via an (α, n) reaction which requires a source of high energy alpha particles [2,26,27], such as the 9 Be(α, n) 12 C reaction in beryllium. In particular, polonium-210 has been frequently used as the alpha source for such applications due to its high alpha emission energy of 5.4 MeV [30,31], as well as a high speci c activity of 4 kCi/g, which stems from its relatively low atomic mass and short half-life of 138 days [3]. ...
... Thus, a chain reaction can be initiated in a mixture of atoms of bismuth-209 and light elements, using an initial neutron source to transmute the bismuth atoms into that of polonium, which would themselves emit high-energy alpha particles that impact the light elements, causing further neutron production. Furthermore, 209 Bi and 9 Be are also able to cause neutron multiplication via a (n, 2n) reaction with fast neutrons [6], increasing the neutron population of the system. To be suitable for sustaining neutron multiplication, the bismuth salt should contain a light element with high neutron yields from (α, n) reactions [7], e.g. ...
... To be suitable for sustaining neutron multiplication, the bismuth salt should contain a light element with high neutron yields from (α, n) reactions [7], e.g. 27 Al, 19 F, 10 B, 9 Be, as well as a high hydrogen content for neutron moderation, as the 209 Bi neutron capture cross section for 210 Po formation is greatly increased at thermal neutron velocities. Hence, bismuth beryllium acetate was used due to its high beryllium content, which is ideal for alpha-to-neutron conversion, as well as its intrinsic neutron multiplication and moderation properties. ...
Neutronic Chain Reactions for Polonium-210 Production
Preprint
Full-text available
  • Oct 2023
The production of the industrially significant radionuclide polonium-210 from the neutron irradiation of bismuth metal and the subsequent beta decay of bismuth-210 is highly inefficient due to the small neutron capture cross section of bismuth-209. In this paper, we report a previously undescribed self-sustaining nuclear chain reaction involving self-propagating neutron multiplication in bismuth salts that allow for rapid and cost-effective production of polonium-210. The reaction proceeds in a cycle of three alternating elementary steps – the capture of neutrons by bismuth-209 and the subsequent formation of polonium-210, the emission of high-energy alpha particles by polonium-210, and the production of more neutrons from (α,n) and (n,2n) reactions on light element and bismuth-209 nuclei respectively. Furthermore, the high hydrogen density of the compound also confers it intrinsic neutron moderation properties, increasing the neutron capture cross section of bismuth-209 at thermal neutron energies. The chain reaction was proven to have successfully occurred by irradiating a sample of the bismuth salt with a 80 µCi neutron source and monitoring the activity levels of the reaction. It was found that the activity of the reaction increased exponentially after an initial stable period following a derived formula for polonium production trends for the reaction, thus validating the occurrence of the reaction. Furthermore, alpha spectroscopy confirmed that polonium-210 had been produced by characterising the 5.30 MeV alpha emission peak of the reaction in addition to using beta spectroscopy to identify the parent nuclide bismuth-210, further proving that the reaction was successful. Hence, this paper reports the successful initiation and characterisation of a novel nuclear chain reaction, and its potential applications offered by a method of rapidly producing large quantities of polonium-210.
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... Since then, this method of neutron generation by irradiating light elements such as beryllium, lithium, and fluorine with an intense alpha source has been incorporated into radioisotopic neutron sources used in nuclear laboratories throughout the world. The mechanism by which the neutrons are produced proceeds via an (α, n) reaction which requires a source of high energy alpha particles [2,26,27], such as the 9 Be(α, n) 12 C reaction in beryllium. In particular, polonium-210 has been frequently used as the alpha source for such applications due to its high alpha emission energy of 5.4 MeV [30,31], as well as a high specific activity of 4 kCi/g, which stems from its relatively low atomic mass and short half-life of 138 days [3]. ...
... Thus, a chain reaction can be initiated in a mixture of atoms of bismuth-209 and light elements, using an initial neutron source to transmute the bismuth atoms into that of polonium, which would themselves emit high-energy alpha particles that impact the light elements, causing further neutron production. Furthermore, 209 Bi and 9 Be are also able to cause neutron multiplication via a (n, 2n) reaction with fast neutrons [6], increasing the neutron population of the system. ...
... To be suitable for sustaining neutron multiplication, the bismuth salt should contain a light element with high neutron yields from (α, n) reactions [7], e.g. 27 Al, 19 F, 10 B, 9 Be, as well as a high hydrogen content for neutron moderation, as the 209 Bi neutron capture cross section for 210 Po formation is greatly increased at thermal neutron velocities. Hence, bismuth beryllium acetate was used due to its high beryllium content, which is ideal for alpha-to-neutron conversion, as well as its intrinsic neutron multiplication and moderation properties. ...
Neutronic Chain Reactions in Bismuth Salts
Preprint
  • Jun 2023
The production of the industrially significant radionuclide polonium-210 from the neutron irradiation of bismuth metal and the subsequent beta decay of bismuth-210 is highly inefficient due to the small neutron capture cross section of bismuth-209. In this paper, we report a previously undescribed self-sustaining nuclear chain reaction involving self-propagating neutron multiplication in bismuth salts that allow for rapid and cost-effective production of polonium- 210. The reaction proceeds in a cycle of three alternating elementary steps - the capture of neutrons by bismuth-209 and the subsequent formation of polonium-210, the emission of high-energy alpha particles by polonium-210, and the production of more neutrons from (alpha,n) and (n,2n) reactions on light element and bismuth-209 nuclei respectively. The chain reaction was proven to have successfully occurred by irradiating a sample of the bismuth salt with a 80 microcurie neutron source and monitoring the activity levels of the reaction. It was found that the activity of the reaction increased exponentially after an initial stable period following a derived formula for polonium production trends for the reaction, thus validating the occurrence of the reaction. Furthermore, alpha spectroscopy confirmed that polonium-210 had been produced by characterising the 5.30 MeV alpha emission peak of the reaction in addition to using beta spectroscopy to identify the parent nuclide bismuth-210, further proving that the reaction was successful.rge quantities of polonium-210.
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... Since then, this method of neutron generation by irradiating light elements such as beryllium, lithium, and fluorine with an intense alpha source has been incorporated into radioisotopic neutron sources used in nuclear laboratories throughout the world. The mechanism by which the neutrons are produced proceeds via an (α, n) reaction which requires a source of high energy alpha particles [2,26,27], such as the 9 Be(α, n) 12 C reaction in beryllium. In particular, polonium-210 has been frequently used as the alpha source for such applications due to its high alpha emission energy of 5.4 MeV [30,31], as well as a high specific activity of 4 kCi/g, which stems from its relatively low atomic mass and short half-life of 138 days [3]. ...
... Thus, a chain reaction can be initiated in a mixture of atoms of bismuth-209 and light elements, using an initial neutron source to transmute the bismuth atoms into that of polonium, which would themselves emit high-energy alpha particles that impact the light elements, causing further neutron production. Furthermore, 209 Bi and 9 Be are also able to cause neutron multiplication via a (n, 2n) reaction with fast neutrons [6], increasing the neutron population of the system. ...
... To be suitable for sustaining neutron multiplication, the bismuth salt should contain a light element with high neutron yields from (α, n) reactions [7], e.g. 27 Al, 19 F, 10 B, 9 Be, as well as a high hydrogen content for neutron moderation, as the 209 Bi neutron capture cross section for 210 Po formation is greatly increased at thermal neutron velocities. Hence, bismuth beryllium acetate was used due to its high beryllium content, which is ideal for alpha-to-neutron conversion, as well as its intrinsic neutron multiplication and moderation properties. ...
Neutronic Chain Reactions In Bismuth Salts
Preprint
Full-text available
  • Jun 2023
The production of the industrially significant radionuclide polonium-210 from the neutron irradiation of bismuth metal and the subsequent beta decay of bismuth-210 is highly inefficient due to the small neutron capture cross section of bismuth-209. In this paper, we report a previously undescribed self-sustaining nuclear chain reaction involving self-propagating neutron multiplication in bismuth salts that allow for rapid and cost-effective production of polonium-210. The reaction proceeds in a cycle of three alternating elementary steps-the capture of neutrons by bismuth-209 and the subsequent formation of polonium-210, the emission of high-energy alpha particles by polonium-210, and the production of more neutrons from (α,n) and (n,2n) reactions on light element and bismuth-209 nuclei respectively. Furthermore, the high hydrogen density of the compound also confers it intrinsic neutron moderation properties, increasing the neutron capture cross section of bismuth-209 at thermal neutron energies. The chain reaction was proven to have successfully occurred by irradiating a sample of the bismuth salt with a 80 μCi neutron source and monitoring the activity levels of the reaction. It was found that the activity of the reaction increased exponentially after an initial stable period following a derived formula for polonium production trends for the reaction, thus validating the occurrence of the reaction. Furthermore, alpha spectroscopy confirmed that polonium-210 had been produced by characterising the 5.30 MeV alpha emission peak of the reaction in addition to using beta spectroscopy to identify the parent nuclide bismuth-210, further proving that the reaction was successful. Hence, this paper reports the successful initiation and characterisation of a novel nuclear chain reaction, and its potential applications offered by a method of rapidly producing large quantities of polonium-210.
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... In the series of elements detection in bulk samples by radio isotopic-based PGNAA technique, a wide range of PGNAA geometries employing 252 Cf and 241 Am-Be sources have been reported (Zhang and Tuo, 2014;Anderson et al., 2016;Waheed et al., 2017;Idiri et al., 2010). However, in this work 252 Cf is preferred due to its well-defined energy spectrum and higher specific activity compared to 241 Am-Be source. ...
... The moderating materials should be of large scattering cross section, small absorption cross section and large energy loss per collision (Waheed et al., 2017). Among several materials, HDPE (C 2 H 4 ), paraffin wax (C n H 2n+2 ), and light water (H 2 O) have been commonly used as fast neutron moderators for 252 Cf-based PGNAA geometries and shown to have excellent fast neutron moderating capabilities down to thermal energies (Hadad et al., 2016;Zhang and Tuo, 2014;Wang et al., 2011;Waheed et al., 2017). ...
... The moderating materials should be of large scattering cross section, small absorption cross section and large energy loss per collision (Waheed et al., 2017). Among several materials, HDPE (C 2 H 4 ), paraffin wax (C n H 2n+2 ), and light water (H 2 O) have been commonly used as fast neutron moderators for 252 Cf-based PGNAA geometries and shown to have excellent fast neutron moderating capabilities down to thermal energies (Hadad et al., 2016;Zhang and Tuo, 2014;Wang et al., 2011;Waheed et al., 2017). However, fast neutrons may be slowed to a E-mail address: nelsheikh@bu.edu.sa. ...
Characterization of (252Cf-ZrH2) Monte Carlo model for detection of nitrogen and chlorine by thermal neutron-capture PGNAA
Article
  • Jun 2021
  • RADIAT PHYS CHEM
Monte Carlo calculations were conducted to explore the potentials of zirconium hydride (ZrH2) as possible alternative moderator for high density polyethylene (HDPE) and water (H2O) in a simple (100MBq) ²⁵²Cf-based Prompt Gamma Neutron Activation Analysis (PGNAA) assembly. The thickness of moderators was optimized against thermal neutron flux (φth) as well as the ratio of (thermal/fast) neutron flux (φth/φfast), calculated at the sample cavity. The capabilities of moderating ²⁵²Cf source neutrons and minimizing the γ-background due to ²⁵²⁵Cf fission γ-rays, ¹H(nth,γ)²H γ-rays originates in the moderator as well as other (nth,γ) capture reactions from surroundings were investigated. The optimal selected moderator with sufficient (φth) at the sample cavity and minimum γ-contamination was selected as ZrH2 for use in a ²⁵²Cf-PGNAA for detection of nitrogen and chlorine in bulk sample simulants of melamine (C3H6N6) and sodium chloride (NaCl). The prompt γ-rays spectra were analyzed to detect the 10.83MeV and 6.11 MeV prompt γ-peaks released by ¹⁴N(nth,γ)¹⁵N and ³⁵Cl(nth,γ)³⁶Cl, respectively. The results showed that the (²⁵²Cf-ZrH2)-PGNAA model has positively identified nitrogen with net counts (1.25×10⁵ n.cm⁻².s⁻¹) and Chlorine with net counts of (4.19×10⁶ n.cm⁻².s⁻¹), achieving minimum detectable concentrations of about (350 ppm at 10.83 MeV) and (640 ppm at 6.11 MeV), respectively.
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... The fast neutrons so produced are then moderated using high density polyethylene (HDPE) sheets. The variation of thermal neutron flux with thickness of the moderator sheet is plotted in fig 2. It can be seen that with increase in the moderator thickness, the flux of thermal neutrons increases and reaches maximum for a thickness of 9 cm which is in good agreement with the measurements reported in the literature [25,26]. ...
Thickness Dependent Sensitivity of GAGG:Ce Scintillation detectors for Thermal Neutrons: GEANT4 Simulations and Experimental Measurements
Preprint
  • Jun 2023
In the present work, we report extensive GEANT4 simulations in order to study the dependence of sensitivity of GAGG:Ce scintillation crystal based detector on thickness of the crystal. All the simulations are made considering a thermalised Am-Be neutron source. The simulations are validated, qualitatively and quantitatively, by comparing the simulated energy spectra and sensitivity values with those obtained from experimental measurements carried out using two different thicknesses of the crystal from our own experiment (0.5mm and 3mm) and validated with three other thicknesses (0.01mm, 0.1 mm and 1 mm) from literature. In this study, we define sensitivity of GAGG:Ce as the ratio of area under 77 keV sum peak to 45 keV peak. The present studies clearly confirm that, while it requires about 0.1 mm thickness for the GAGG:Ce crystal to fully absorb thermal neutrons, it requires about 3 mm to fully absorb the thermal neutron induced events. Further, we propose an equation, that can be used to estimate the thickness of the GAGG:Ce crystal directly from the observed sensitivity of the GAGG:Ce crystal. This equation could be very useful for the neutron imaging community for medical and space applications, as well as for manufactures of cameras meant for nuclear security purposes.
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... -2 s -1 for 250 kW [32]. The thermal neutron flux was flutter due to the reactor power noise on the neutron distribution that had effect on random neutron fluctuation [35]. The epithermal neutron flux and thermal neutron flux of Phase 3 and Phase 2 has been displayed on the Table 4.4. ...
A significant review on beam optimization for boron neutron capture therapy A significant review on beam optimization for boron neutron capture therapy
Article
  • Mar 2020
The thermal column at the TRIGA PUSPATI (RTP) research reactor can produce thermal neutron. However, the optimization on the thermal neutron flux produced should be performed to gain a sufficient thermal neutron for boron neutron capture therapy purpose. Thus, the objective of this review is to optimize the thermal neutron flux by designing the collimator with different materials at the thermal column. In order to fulfil the requirement, set by the IAEA standard, the study of BNCT around the world was being reviewed to study the suitable measurement, material, design, and modification for BNCT at the thermal column of TRIGA MARK-II, Malaysia. Initially, the BNCT mechanisms and history was review. Then, this paper review on the design and modifications for BNCT purpose around the world. Based on this review, suitable material and design can be used for the BNCT in Malaysia. Moreover, this paper also reviews the current status of BNCT at the RTP with the measurement of the thermal neutron flux was conducted along the thermal column at 250 kW. The thermal column of RTP was divided into 3 phases (Phase 1, Phase 2 and Phase 3) so that an accurate measurement can be obtained by using gold foil activation method. This value was used as a benchmark for the neutron flux produced from the thermal column. The collimator was designed using different types of materials, and their characteristic towards gamma and neutron flux was investigated. The reviewed demonstrated that the final thermal neutron flux produced was significantly for BNCT purpose. Lastly, this paper recommends the future research can be conducted on BNCT at RTP.
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Optimization in nuclear systems
Chapter
  • Jan 2022
This chapter presents applications of optimization in nuclear systems in the domains of nuclear fission reactors, nuclear fusion reactors, and non-power nuclear systems and applications. The underlying mathematics of the optimization results presented here have been covered in Chapter 9. For nuclear fission reactors, optimization applications in reactor core design, radiation shielding, fuel reloading, and control systems are described. For nuclear fusion reactors, the neutronics optimization of tritium breeding and radiation shielding are described. In non-power applications, the topics reviewed are the optimization of a BF3 detector and a thermal neutron activation analysis detection system.
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Research and applications of nuclear tracks: Developments in Pakistan and global comparison
Article
Full-text available
  • Jan 2012
The present paper describes the development and applications of nuclear track detection technique in Pakistan. Pakistan entered in the field of nuclear tracks in early 1970s when it was still quite new. Highlights of successes of different Pakistani laboratories, working on nuclear tracks, achieved on their own or in collaboration with similar centers in the world are described briefly. The robust features of this investigation are the comprehensive investigation of the addressed research, analysis and review of results, and discussions with the perspectives of applications and new research directions. Further analysis of the published results by the present author and some new results are also presented. This paper portrays a comprehensive picture of the nuclear track detection research and technology in Pakistan and can be useful for a similar development in any country around the globe.
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NEUTRON CAPTURE CROSS SECTION MEASUREMENTS OF PARAFFIN WAX
Article
Full-text available
  • Apr 2014
A solid state track detector, SSNTD, has been used to determine the neutron capture cross-section for paraffin wax. 200 CR-39 detectors (1cm*1cm) are distributed at different height of paraffin wax, which are irradiated using a 38 mCi Am-Be neutron source. It was found the neutron capture cross-section is (σc=0. 624 ±6.5 %), the number of neutron collisions leads to the thermal thermal energy stage is about 21 Collisions at depth reached to 26 cm. From these results one can conclude that this material at limited dimension is more applicable, as a shielding material with neutron sources, in slowing down neutrons from the fast energy energy stage and can define the neutron energy at a certain depth of the wax material.
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Thermal neutron detection using a silicon pad detector and (LiF)-Li-6 removable converters
Article
Full-text available
  • Mar 2013
A semiconductor detector coupled with a neutron converter is a good candidate for neutron detection, especially for its compactness and reliability if compared with other devices, such as 3He tubes, even though its intrinsic efficiency is rather lower. In this paper we show a neutron detector design consisting of a 3 cm × 3 cm silicon pad detector coupled with one or two external 6LiF layers, enriched in 6Li at 95%, placed in contact with the Si active surfaces. This prototype, first characterized and tested at INFN Laboratori Nazionali del Sud and then at JRC Ispra, was successfully shown to detect thermal neutrons with the expected efficiency and an outstanding gamma rejection capability.
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1. Studies on optimization of Moderator thickness for BF3 detectors used for monitoring of fissile material
Article
Full-text available
  • Nov 2012
Monitoring of fissile material content requires the energy independent response due to inherent moderation in waste matrix present inside the waste drum. The detector assembly consists of three BF3 detectors each being 100 cm long and covered with 3.7 cm thick HDPE moderators arranged in a linear geometry. The response of the detector assembly as measured with neutron sources of different energies was observed to be consistent within acceptable limits. Response of the detector assembly for Cf252 source which has a fission neutron spectrum similar to that of fissile material was obtained and compared with FLUKA based simulation. The sensitivity of the system as evaluated using FLUKA simulation for a Cf-252 neutron source was found to be in good agreement with the experimental values obtained. This study indicates that the optimum moderator thickness for effective monitoring of fissile material using BF3 neutron detector assembly is 6-7 cm. Keywords: Fissile material, linear geometry, neutron monitoring, BF3 detector
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Development of a spherical tissue equivalent proportional counter for neutron monitoring
Article
  • Dec 2015
A spherical tissue equivalent proportional counter (TEPC) for neutron monitoring has been developed. It was properly designed to produce a uniform electric field intensity around the anode wire. An internal 241Am alpha source was adopted for lineal energy calibration. The TEPC was characterized in terms of dose equivalent response in a standard 252Cf neutron field, and was tested with 2.45 MeV neutrons. Microdosimetric spectra, frequency mean lineal energy and dose-average mean lineal energy of 2.45 MeV neutrons were obtained and compared with FLUKA Monte Carlo simulation results. The measurement and simulation results agreed well. The mean quality factor and dose equivalent values evaluated from the 2.45 MeV neutron measurement were in good agreement with the recommended effective quality factor and ambient dose equivalent H∗ (10), respectively. Preliminary results have proved the availability of the developed TEPC for neutron monitoring.
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Energy spectrum measurement and dose rate estimation of natural neutrons in Tibet region
Article
  • Dec 2015
In this work, natural neutron spectra at nine sites in Tibet region were measured using a multi-sphere neutron spectrometer. The altitude-dependence of the spectra total fluence rate and ambient dose equivalent rate were analyzed. From the normalized natural neutron spectra at different altitudes, the spectrum fractions for neutrons of greater than 0.1 MeV do not differ obviously, while those of the thermal neutrons differ greatly from each other. The total fluence rate, effective dose rate and the ambient dose equivalent rate varied with the altitude according to an exponential law.
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Investigation of the neutron spectrum of americium–beryllium sources by Bonner sphere spectrometry
Article
  • Nov 2014
  • NUCL INSTRUM METH A
Americium beryllium neutron sources are certainly the most widely used in neutron dosimetry laboratories, basically due to their long half-life and their energy distribution, which covers the energy domain of interest for many applications in ambient and personal dosimetry. Nevertheless, the spectrum of this source depends on the materials and dimension of the capsule and on the amount and physical chemical properties of the active material, thus affecting relevant quantities such as the spectrum averaged fluence-to-dose equivalent conversion coefficient. A EURAMET (European Association of National Metrology Institutes) project (n. 1104) was initiated to experimentally investigate how the neutron spectrum changes for different source sizes and encapsulations with a view to providing improved data for a planned revision of the ISO 8529 Standard Series. The experimental campaign was carried out in the low scatter facility at NPL. Here three different Bonner sphere spectrometers, BSSs, vvere exposed to the neutron fields produced by three different neutron sources formats: one X3 capsule (1 Ci) and two X14 capsules (10 Ci and 15 Ci). The specific advantage of the BSS is the large sensitivity to low energy neutrons (E < 0.1 MeV) which is the component expected to be most affected by the capsule-to-capsule variations and the component which is least well known. This paper summarises the results of the campaign with emphasis on (1) estimating the low-energy component or the Am-Be neutron spectrum, according to the encapsulation type; (2) evaluating the coherence between the Bonner spheres data and the previous studies performed with high-resolution spectrometers but limited in energy to E > 0.1 MeV; (3) understanding whether the ISO-recommended Am-Be spectrum needs to be amended, and for which source formats.
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Optimizing moderator dimensions for neutron scattering at the spallation neutron source
Article
  • Dec 2013
  • REV SCI INSTRUM
In this work, we investigate the effect of neutron moderator dimensions on the performance of neutron scattering instruments at the Spallation Neutron Source (SNS). In a recent study of the planned second target station at the SNS facility, we have found that the dimensions of a moderator play a significant role in determining its surface brightness. A smaller moderator may be significantly brighter over a smaller viewing area. One of the immediate implications of this finding is that for modern neutron scattering instrument designs, moderator dimensions and brightness have to be incorporated as an integrated optimization parameter. Here, we establish a strategy of matching neutron scattering instruments with moderators using analytical and Monte Carlo techniques. In order to simplify our treatment, we group the instruments into two broad categories: those with natural collimation and those that use neutron guide systems. For instruments using natural collimation, the optimal moderator selection depends on the size of the moderator, the sample, and the moderator brightness. The desired beam divergence only plays a role in determining the distance between sample and moderator. For instruments using neutron optical systems, the smallest moderator available that is larger than the entrance dimension of the closest optical element will perform the best (assuming, as is the case here that smaller moderators are brighter).
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High resolution measurements of neutron energy spectra from AmBe and AmB neutron sources
Article
  • Dec 1995
  • NUCL INSTRUM METH A
A 3He sandwich spectrometer incorporating two semiconductor detectors and a proportional counter region has been designed and constructed to perform high resolution measurements of neutron energy spectra in the energy range 100 keV to 15 MeV. The efficiency of the spectrometer was determined experimentally, using a 252Cf spontaneous fission source, in the low-scatter facility of the National Physical Laboratory. The Monte Carlo technique was also used to determine the efficiency and these calculated values were used to extrapolate the measured efficiency to higher energies. The neutron energy spectra from two different sized AmBe neutron sources and an AmB neutron source were measured. Although these spectra have been measured previously, the present work represents an improvement in the energy resolution and extends the energy range. Average values for dose equivalent quantities for neutrons from these sources are also presented. The spectral data is available from the authors, as a printed table or on a computer diskette, on request.
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